[Federal Register Volume 59, Number 138 (Wednesday, July 20, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-17503]
[[Page Unknown]]
[Federal Register: July 20, 1994]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from June 24, 1994, through July 8, 1994. The
last biweekly notice was published on July 6, 1994 (59 FR 34657).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, and should cite the publication date and page
number of this Federal Register notice. Written comments may also be
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the NRC Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC
20555. The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By August 19, 1994, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC 20555 and at the local
public document room for the particular facility involved. If a request
for a hearing or petition for leave to intervene is filed by the above
date, the Commission or an Atomic Safety and Licensing Board,
designated by the Commission or by the Chairman of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the designated Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington, DC 20555, by the
above date. Where petitions are filed during the last 10 days of the
notice period, it is requested that the petitioner promptly so inform
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and
to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1) (i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC 20555, and at the local public document
room for the particular facility involved.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station (PVNGS),
Unit Nos. 1, 2, and 3, Maricopa County, Arizona
Date of amendment requests: June 17, 1994.
Description of amendment requests: The proposed changes would
enhance the PVNGS technical specifications (TS) by removing five tables
of component lists in accordance with NRC Generic Letter (GL) 91-08,
``Removal of Component Lists from Technical Specifications.'' The
affected tables are 3.3-9B, 3.3-9C, 3.6-1, 3.8-2, and 3.8-3. The
references to these five tables will also be removed from the text of
the TS in accordance with the sample TS change amendment provided by
the NRC in GL 91-08. These five removed tables will be incorporated
into a new document, which will be administratively controlled
according to the change controls provisions of the TS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis about the issue of no significant hazards
consideration, which is presented below:
Standard 1--Involve a significant increase in the probability or
consequence of an accident previously evaluated:
The proposed amendment will remove five tables of component lists
from the TS and add them to an administratively controlled document
which is subject to the change controls provisions of the TS. The new
location of the affected component lists is easily retrievable. The
existing TS requirements and those components to which they apply are
not altered by this TS amendment. There are no changes to the
operations, maintenance, surveillance, and/or qualification of any
component on the removed lists. Therefore, the probability of
occurrence and the consequences of any previously evaluated accident is
[sic] not changed.
Standard 2--Create the possibility of a new or different kind of
accident from any accident previously evaluated:
The TS requirements and the components to which they apply are not
altered by this amendment. The removed component lists are added to a
controlled and easily retrievable document. This amendment has no
impact on plant operations, maintenance, testing, or component
qualification. Therefore, the possibility of a new or different kind of
accident is not created by this amendment.
Standard 3--Involve a significant reduction in a margin of safety:
The removal of these five component lists from the TS does not
alter existing TS requirements or those components to which they apply.
These lists will be added to an administratively controlled document
which is subject to the controls provisions of the TS. More
specifically, there is no impact on safe plant shutdown, maintenance or
hot standby, containment isolation capability, containment leakage
rate, and/or the operability of safety related valves. Therefore,
removal of these five component lists from the TS will not involve a
reduction in the margin of safety.
The NRC staff has reviewed the licensees' analysis and, based on
that review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004.
Attorney for licensees: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999.
NRC Project Director: Theodore R. Quay.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County,
Maryland
Date of amendments request: May 27, 1994.
Description of amendments request: The proposed amendment would
revise the Technical Specification surveillance test interval from
monthly to quarterly for several channel functional tests for the
Reactor Protective System and the Engineered Safety Feature Actuation
System (ESFAS). In addition, an administrative change to the ESFAS
table would remove an out-of-date footnote concerning the emergency
diesel generator logic circuit modifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Would not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The Reactor Protective System (RPS) and the Engineered Safety
Features Actuation System (ESFAS) provide the actuation signals to
safety equipment necessary to mitigate design basis accidents and
transients. The proposed change would increase the surveillance test
interval from monthly to quarterly for several of the RPS and ESFAS
instrumentation channel functional tests. The RPS/ESFAS instruments are
not an initiator in any previously evaluated accidents. Therefore, the
proposed changed does not involve an increase in the probability of an
accident previously evaluated. The required plant-specific setpoint
drift analysis for Calvert Cliffs demonstrated that the observed
changes in instrument uncertainties for the extended surveillance test
interval do not exceed the current 30-day setpoint assumptions. This
provides confidence the 90-day test interval will not impact the
ability to detect and monitor system degradation. Therefore, the
proposed change will not change the ability of the RPS/ESFAS
instrumentation to respond to and mitigate the consequences of any
previously evaluated accident. In addition, an obsolete footnote is
removed from ESFAS Table 4.3-2.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. Would not create the possibility of a new or different type of
accident from any accident previously evaluated.
The proposed extended surveillance test interval for the RPS and
ESFAS and the removal of the obsolete footnote does not involve any
changes in equipment or the function of these instruments. The proposed
change does not represent a change in the configuration or operation of
the plant. The RPS and ESFAS setpoints will not be changed as the
instrument uncertainties resulting from the proposed surveillance test
interval (calculated using actual plant data) are less than the
instrument uncertainties assumed for the current surveillance interval.
Therefore, the proposed change does not create the possibility of a new
or different type of accident from any accident previously evaluated.
3. Would not involve a significant reduction in a margin of safety.
The proposed change will not affect the functions of the RPS or the
ESFAS instruments. The CEN-327 and CEN-327, Supplement 1, topical
reports quantified the corresponding changes in core melt frequency for
the representative fault tree models that were developed for Calvert
Cliffs. The proposed change has two principal effects with opposing
impacts on core melt frequency. The first impact is a slight increase
in core melt frequency that results from the increased unavailability
of the instrumentation in question. This assumed unavailability results
from less frequent testing. The unavailability of the tested
instrumentation components represents the potential for the failure of
the reactor to trip, an Anticipated Transient Without Scram, or a
failure of the appropriate engineered safety features to actuate when
required. The opposing impact on core melt risk is the corresponding
reduction in core melt frequency that would result due to the reduced
exposure of the plant to test-induced transients. The two changes are
nearly equal and the net result is no distinguishable effect on plant
safety. The NRC issued a Safety Evaluation Report which found that
these evaluations were acceptable for justifying the extensions in the
surveillance test intervals for the RPS and ESFAS from 30 days to 90
days.
The RPS and ESFAS setpoints will not be changed since the
instrument drift resulting from the proposed surveillance test interval
is less than the instrument drift presently assumed for the current
surveillance interval. This provides confidence the 90-day test
interval will not impact the ability to detect and monitor system
degradation. The removal of the ESFAS Table footnote only removes
obsolete information from the Technical Specifications. The conclusions
of the accident analyses in the Calvert Cliffs Updated Safety Analysis
Report remain valid and the safety limits continue to be met.
Therefore, the proposed change does not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involves no significant hazards consideration.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Michael L. Boyle.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County,
Maryland
Date of amendments request: June 8, 1994.
Description of amendments request: The proposed amendments would
revise the Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 (CC1/
2), Technical Specifications (TS) 4.6.2.2.b to extend the surveillance
interval for the containment fan coolers from 18 months to 24 months.
The requested changes have been submitted in accordance with Generic
Letter (GL) 91-04, ``Guidance on Preparation of a License Amendment
Request for Changes in Surveillance Intervals to Accommodate a 24-Month
Fuel Cycle.'' These proposed amendments are part of a series of
requests that will eliminate the need for mid-cycle surveillance
outages to accommodate the existing 18 month surveillance requirements
since CC1-2 is operating on 24-month fuel cycles.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Would not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The purpose of the Containment Air Cooling (CAC) System is to cool
the containment atmosphere, and thereby limit containment pressure and
temperature, following a Loss of Coolant Accident (LOCA) or Main Stream
Line Break in containment. Failure of the CAC System is not an
initiator for any previously analyzed accident. Therefore, the proposed
change does not involve an increase in the probability of an accident
previously evaluated.
Historical CAC System reliability, monthly surveillances and
monitoring of CAC-related plant parameters provide assurance that
undetected system degradation will not occur between 24-month
surveillances, and the system will continue to perform its safety
function. Therefore, there will be no significant increase in the
consequences of accidents previously evaluated. Therefore, the proposed
Technical Specification change does not increase the probability or
consequences of an accident previously evaluated.
2. Would not create the possibility of a new or different type of
accident from any accident previously evaluated.
This requested revision to increase the interval for a CAC
surveillance from 18 to 24 months does not involve a significant change
in the design or operation of the plant. No hardware is being added to
the plant as part of the proposed change. The proposed change will not
introduce any new accident initiators. Therefore, the proposed change
would not create the possibility of a new or different type of accident
from any accident previously evaluated.
3. Does operation of the facility in accordance with the proposed
amendment involve a significant reduction in a margin of safety.
The CAC System provides a margin of safety by providing a means by
which containment pressure can be limited following a LOCA or Main
Steam Line Break. The proposed change does not affect the operation or
design of the CAC System. Historical monthly surveillances and Control
Room indications give assurance that the reduction in surveillance
frequency will not adversely affect our ability to detect degradation
in the system. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involves no significant hazards consideration.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Michael L. Boyle.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County,
Maryland
Date of amendments request: June 8, 1994.
Description of amendments request: The proposed amendments would
revise the Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Technical
Specifications 4.8.1.1.1.b to extend the alternate 69 kV offsite power
circuit surveillance frequency from 18 to 24 months.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Would not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The purpose of the 69 kV Southern Maryland Electric Cooperative
(SMECO) power source is to act as an independent energy source for
achieving and maintaining safe shutdown of the plant if the 500 kV
system is not available. Failure of the 69 kV SMECO system is not an
initiator for any existing accident. Therefore, the proposed change
does not involve an increase in the probability of an accident.
The 69 kV SMECO system could be used to mitigate the consequences
of accidents involving a loss of primary offsite power. However, the
accident analyses assume that if the 500 kV circuits were not
available, the Emergency Diesel Generators would be used to provide
power to maintain the plant in a safe shutdown condition. A historical
review of surveillance test results indicates the system has
experienced only one significant failure in the last ten years. In
addition, the system is routinely used.
However, the SMECO system is not assumed to function in our
accident analysis, so this change will result in no significant
increase in the consequences of accidents previously evaluated.
Therefore, the proposed Technical Specification change does not
increase the probability or consequences of an accident previously
evaluated.
2. Would not create the possibility of a new or different type of
accident from any accident previously evaluated?
This requested increase in the interval for a 69 kV SMECO
surveillance from 18 to 24 months does not involve a significant change
in the design or operation of the plant. No hardware is being added to
the plant of the proposed change. The proposed change will not
introduce any new accident initiators. Therefore, the proposed change
would not create the possibility of a new or different type of accident
from any accident previously evaluated.
3. Does operation of the facility in accordance with the proposed
amendment involve a significant reduction in a margin of safety?
The 69 kV SMECO system provides a margin of safety by providing an
alternate offsite electrical power source. The proposed change does not
affect the operation or design of the 69 kV SMECO system. Historical
surveillance data and routine use indicates that the reduction in
surveillance frequency will not adversely affect our ability to detect
degradation in the system. Therefore, the proposed change does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involves no significant hazards consideration.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Michael L. Boyle
Baltimore Gas and Electric Company, Docket No. 50-318, Calvert Cliffs
Nuclear Power Plant, Unit No. 2, Calvert County, Maryland
Date of amendment request: May 27, 1994.
Description of amendment request: This amendment would revise the
existing heatup and cooldown curves and rates to increase their
applicability to 30 Effective Full-Power Years (EFPY). The fluence
value that was used to determine the heatup and cooldown curves was
based on the peak fluence and EFPY at the end of Cycle 9 and the peak
predicted fluence for Cycles 10 and beyond. In addition, a variable-
setpoint low-temperature overpressure protection (VLTOP) system is
being installed to increase the allowable operating pressure band in
the Low-Temperature Overpressure Protection (LTOP) region. The VLTOP
system uses a variable power-operated relief valve setpoint to take
advantage of increased Appendix G pressure limits at increased reactor
coolant system (RCS) temperatures. This system will increase the
operating window in which the plant may operate during heatup and
cooldown.
Specifically, this amendment would revise the Unit No. 2 heatup and
cooldown curves and rates for the following Technical Specification
(TS) sections. TS Section 3.4.9.1.a would be revised to decrease the
maximum heatup rate from a fixed value of 75 deg.F/hr to a more
conservative variable heatup rate of 30 deg.F to 60 deg.F/hr which
varies with RCS temperature range from 70 deg.F to greater than 246
deg.F. TS Section 3.4.9.1.b would be revised to increase the RCS
temperatures for the maximum allowable cooldown rates. TS Figures
3.4.9-1 and 3.4.9-2 would be replaced by new RCS pressure Temperature
Limits. The revised curves and rates are based on the predicted fluence
value for Cycle 10 and beyond.
The following TS sections would be revised to support modifications
to the LTOP system. TS sections 3.4.9.3.a.1 and 3.4.9.3.a.2 would be
changed to ``trip setpoint below the curve in Figure 3.4.9-3*'' to
account for the VLTOP system. The footnote, ``When on shutdown cooling,
the PORV trip setpoint shall be less than or equal to 443 psia,'' has
been added for Shutdown Cooling Operation to maintain an extra setpoint
that is independent of RCS temperature and is equal to the lowest
variable setpoint. The Minimum Pressure and Temperature (MPT) Enable
would be changed from 305 deg.F to 301 deg.F. This change would
effect TS sections 3.1.2.1, 3.1.2.3, Table 3.3-3, 3.4.1.2, 3.4.1.3,
3.4.3, 3.4.9.3, 4.5.2, and 3.5.3. Due to the lower MPT Enable
temperature, the transition region at which the high pressure safety
injection pumps are placed under manual control on cooldown and
restored to automatic status on heatup would be changed from a
temperatue range of 305 deg.F-350 deg.F, to 301 deg.F-325 deg.F.
This affects TS 3.5.3 and Table 3.3-3. TS Bases Sections B3/4.4.1, B3/
4.4.9, and 3/4.5.2 would change to be consistent with the proposed
change and to provide additional clarification of some of the existing
bases.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Would not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The Low Temperature Overpressure Protection (LTOP) system,
including the administrative controls, ensures that the 10 CFR Part 50,
Appendix G, Pressure-Temperature (P-T) limits for the reactor pressure
vessel will not be violated while operating at low temperatures. The
heatup and cooldown curves are conservatively developed in accordance
with the fracture toughness requirements of 10 CFR Part 50, Appendix G,
as supplemented by the American Society of Mechanical Engineers Boiler
and Pressure Vessel Code Section III, Appendix G. The reactor vessel
material Adjusted RTNDT values are based on the conservative
methodology provided in Regulatory Guide 1.99, Revision 2.
Analyses show that the proposed use of a variable LTOP system will
not result in a significant increase in the probability of an
inadvertent opening of a Power-Operated Relief Valve (PORV) causing a
small break Loss-of-Coolant-Accident. The proposed heatup and cooldown
curves and associated limits continue to provide conservative
restrictions on Reactor Coolant System (RCS) pressure to minimize
material stresses in the RCS due to normal operating transients, thus
minimizing the likelihood of a rapidly propagating fracture due to
pressure transients at low temperatures. Because the proposed heatup
and cooldown curves and rates are based on conservative Appendix G
methods, and because the LTOP controls protect the Appendix G P-T
limits, the proposed curves and limits do not involve an increase in
the probability of accidents previously evaluated.
The proposed use of a variable PORV trip setpoint and the increase
in the allowable fluence at the reactor vessel wall results in the
changes to the heatup and cooldown curves and rates, the Minimum
Pressure and Temperature (MPT) Enable temperature, and high pressure
safety injection pump manual control transition temperature. These
proposed changes continue to provide sufficient margin to accommodate
postulated pressurization from mass and energy addition transients.
Calculations have been performed that predict the response to such
transients. Because the results of the analyses remain well within the
conservative acceptance limits of Appendix G, these changes do not
increase the consequences of accidents previously evaluated.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. Would not create the possibility of a new or different type of
accident from any accident previously evaluated.
The new variable LTOP control system along with the proposed
changes to the Technical Specifications will ensure that the Appendix G
P-T limits will not be violated during low temperature operations.
While setpoints and curves have changed, this proposed change does not
introduce any operator actions that are significantly different from
current operator actions used at the plant. The variable LTOP system
will continue to have redundant channels to ensure that no single
equipment failure or operator error will result in violation of the P-T
limits. The use of a variable LTOP system does not create a new failure
mechanism for the PORV. The failure mechanism for the PORV continues to
be an inadvertent opening or the failure to open during a pressure
transient which has been previously evaluated. Therefore, the proposed
change does not create the possibility of a new or different type of
accident from any accident previously evaluated.
3. Would not involve a significant reduction in a margin of safety.
This change will ensure that the margin of safety is maintained
with respect to energy or mass addition events in that none of the
events postulated could challenge the Appendix G limits. The proposed
use of a variable PORV trip setpoint and the increase in the allowable
fluence at the reactor vessel wall necessitate the changes to the
heatup and cooldown curves and rates, the MPT Enable temperature, and
high pressure safety injection pump manual control transition
temperature. These changes ensure that the margin of safety is
maintained by protecting the Appendix G limits for all postulated
transients. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Michael L. Boyle.
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of amendments request: June 17, 1994.
Description of amendments request: The proposed amendments would
revise the Technical Specifications (TS) to (1) remove the heatup and
cooldown curves from TS 3/4.4.6 and relocate them to a newly created
Pressure and Temperature Limits Report, and (2) remove the reactor
vessel material surveillance program withdrawal schedule from TS Table
4.4.6.3-1 and relocate it to the Updated Final Safety Analysis Report.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment [sic] does not involve a significant
increase in the probability or consequences of an accident previously
evaluated because the changes are administrative in nature. These
changes do not alter the configuration or operation of the facility.
The Limiting Safety Systems Settings and Safety Limits specified in the
current Technical Specifications remain unchanged.
2. The proposed amendment [sic] does not create the possibility of
a new or different kind of accident from any accident previously
evaluated. The safety analysis of the facility remains complete and
accurate. There are no physical changes to the facility and the plant
conditions for which the design basis accidents have been evaluated are
still valid. The operating procedures and emergency procedures are
unaffected.
3. The proposed amendment [sic] does not involve a significant
reduction in the margin of safety because these margins are established
through the Limiting Conditions of Operation, Limiting Safety System
Settings and Safety Limits specified in the Technical Specifications,
and since there are no changes to the physical design or operation of
the facility, these margins will not be changed.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602.
NRC Project Director: William H. Bateman.
Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station, Units 1 and 2, Lake County, Illinois
Date of amendment request: January 19, 1994.
Description of amendment request: The proposed amendment would
revise the Technical Specifications by increasing the minimum reactor
coolant system temperature required for criticality from 500 degrees
Fahrenheit to 530 degrees Fahrenheit.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed change does not result in a significant increase in
the probability or consequences of accidents previously evaluated. The
probability for an accident is independent of the changes being
proposed. Reactor criticality at 530 degrees Fahrenheit instead of the
nominal no-load Tavg of 547 degrees Fahrenheit does not affect any
of the accident initiators in the analyses, but does change one of the
initial conditions assumed in the Safety Analysis. However, the change
in initial conditions from the nominal no-load temperature of 547
degrees Fahrenheit to 530 degrees Fahrenheit does not increase the
probability of any of the events considered in the Safety Analysis. The
proposed Minimum Temperature for Criticality specification (530 degrees
Fahrenheit) will be more restrictive than the current specification
which allows reactor criticality at a temperature as low as 500 degrees
Fahrenheit. In addition, the Action Statement will require operator
response to place the reactor in a subcritical condition (Mode 3) with
15 minutes should the temperature drop below the limit for greater than
a specified amount of time (15 minutes).
Likewise, the proposed change does not significantly increase the
consequences of an accident previously evaluated. In the reanalysis of
the Zero Power accidents (Rod Withdrawal From Subcritical, Rod
Ejection, Main Steamline Break, Boron Dilution During Startup, and
Feedwater Malfunction) from an initial condition of 530 degrees
Fahrenheit, it was concluded that the results and conclusions in the
current Safety Analysis remain valid based on the fact that the current
analysis results are conservative and bounding for reactor criticality
at 530 degrees Fahrenheit. The LOCA transient analyses are unaffected
by the proposed change since they are initiated from the limiting
condition of 102 percent. Since the full power Tavg value is
unchanged by this proposed amendment, the LOCA analyses are unaffected.
The proposed change will ensure that plant parameters are within their
analyzed ranges prior to reactor criticality and appropriate operator
actions are taken should the temperature drop below the temperature
limit after reaching criticality.
The proposed administrative changes delete requirements which are
no longer applicable and will have no affect on the probability or
consequences of any accident previously evaluated in the analyses.
The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated. The
change does not involve the addition of any new or different type of
equipment, nor does it involve the operation of equipment required for
safe operation of the facility in a manner different from those
addressed in the Final Safety Analysis Report. The proposed change will
ensure that plant parameters are within their analyzed ranges prior to
reactor criticality. The proposed administrative changes delete
requirements which are no longer applicable and will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed change does not involve a significant reduction in a
margin of safety. The proposed change does not affect any safety
related system or component operation or operability, instrument
operation, or safety system setpoints, and does not result in increased
severity of any of the accidents considered in the analysis. Operator
response to a drop in temperature after reaching criticality for a
specified period of time (greater than 15 minutes) will place the
reactor in a subcritical condition which is inherently more stable than
when critical below the Point of Adding Heat. The proposed
administrative changes are being made to clarify Technical
Specifications with no change of intent. Therefore, the proposed
changes do not create a significant reduction in a margin of safety.
In conclusion, based on the previous considerations, Commonwealth
Edison Company believes that the activities associated with this
Technical Specification amendment request satisfy the Significant
Hazards Consideration standards of 10 CFR 50.92(c) and, accordingly, a
finding that this Technical Specification amendment does not represent
a Significant Hazards Consideration is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Waukegan Public Library, 128
N. County Street, Waukegan, Illinois 60085.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690.
NRC Project Director: Robert A. Capra.
Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station, Units 1 and 2, Lake County, Illinois
Date of amendment request: June 24, 1994.
Description of amendment request: The proposed amendment would
revise the Technical Specifications by removing the containment
recirculation sump level instrumentation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant increase
in the probability of occurrence or consequences of any accident
previously evaluated.
This change does not affect the initiators or precursors of any
accident previously evaluated. This change will not increase the
likelihood that a transient initiating event will occur because
transients are initiated by equipment malfunction and/or catastrophic
system failure. Since the proposed change does not involve the
introduction of new or redesigned plant equipment, failure mechanisms
are not impacted. As a result, the probability of occurrence of
accidents previously evaluated is not increased.
The consequences of accidents previously evaluated are not
increased. Removal of [containment recirculation sump level] CRSL
instrumentation requirements from Technical Specifications does not
affect the ability to mitigate the consequences of any accident
previously evaluated. The CRSL instrumentation is used to verify that
[residual heat removal] RHR pumps have adequate [net positive suction
head] NPSH to operate in the recirculation mode following a Loss of
Coolant Accident (LOCA).
As stated in the Zion Updated Final Safety Analysis Report (UFSAR),
the changeover from the injection mode to the recirculation mode of
emergency core cooling is initiated when the low level alarm on the
RWST annunciates. This occurs when the RWST level drops to 13'-7''
(145,600 gallons).
At this point, sufficient water has been delivered to the
containment, from the Containment Spray (CS) system and, by Emergency
Core Cooling System (ECCS) injection, through the [reactor coolant
system] RCS break, to provide at least one foot of water above the
containment floor. One foot of water above the containment floor
provides sufficient volume to sustain the required NPSH of the RHR
pumps in the recirculation mode of operation. The water level in the
Containment Building during the recirculation phase will be
approximately 5 feet above the floor elevation (568') based on the
volume of the RCS, accumulators, and the minimum required volume of the
RWST. Minimum RWST volume of 350,000 gallons is required by Technical
Specification 3.8.1.F.
In summary, RWST level instrumentation, which satisfies Regulatory
Guide 1.97 qualification requirements for a Type A, Category 1
variable, provides the operator with the primary indication of the
appropriate time to initiate switchover to the recirculation mode, as
well as indication of adequate NPSH for the RHR pumps. Containment
Water Level (wide range) instrumentation which is qualified as a Type B
variable provides confirmatory indication of water level in
containment.
The proposed change does not affect the procedures controlling
operation of equipment, or systems required to mitigate the accidents
considered in the UFSAR. As such, there will be no significant increase
in the consequences of any accident previously evaluated.
2. The proposed amendment does not create the possibility of a new
or different kind of accident from any previously analyzed.
The proposed change does not involve the addition of any new or
different types of equipment, nor does it involve the operation of
equipment required for safe operation of the facility in a manner
different from those addressed in the UFSAR. No safety related
equipment or function will be altered as a result of this proposed
change. Because no new failure modes are introduced, the proposed
amendment does not create a new or different kind of accident from any
previously analyzed in the UFSAR. Also, the methods of recovery from
accidents described in the UFSAR are not affected.
Based on the above discussion, the proposed amendment does not
create a new or different kind of accident from any previously analyzed
in the UFSAR.
3. The proposed changes do not involve a significant reduction in a
margin of safety.
No design margins are impacted and the newly chosen primary
indicator (RWST level) is both consistent with plant emergency
procedures and appropriately qualified. The proposed change will not
adversely impact the peak clad temperature, amount of fuel damage, or
offsite dose projected to occur from the design basis accidents. Thus,
the margin of safety is not diminished.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Waukegan Public Library, 128
N. County Street, Waukegan, Illinois 60085.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690.
NRC Project Director: Robert A. Capra.
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck
Plant, Middlesex County, Connecticut
Date of amendment request: June 16, 1994.
Description of amendment request: The proposed amendment will
remove a footnote applicable for Cycle 18 only regarding the
surveillance of the automatic bus transfer (ABT) system and add
surveillance requirement 4.8.3.1.2, to test the ABT once per refueling.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed changes do not involve an SHC [significant hazards
consideration] because the changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes to delete a note from the limiting condition
for operation for the MCC-5 ABT scheme and to add the surveillance
requirement has no impact on the probability or consequences of an
accident previously evaluated. By removing the requirement to test the
scheme on-line, the probability of failure to mitigate an accident
while the Haddam Neck Plant is operational is incrementally decreased.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed changes remove the requirements to disable the subject
ABT feature for testing, leaving the scheme undisturbed throughout
normal plant operation, and therefore does not create the possibility
of a new accident or different kind of accident from any previously
analyzed.
3. Involve a significant reduction in a margin of safety.
The proposed changes require that the MCC-5 ABT feature be tested
during a plant shutdown rather than during normal operation. This will
place the ABT scheme in a test environment that has no significant
reduction in a margin of safety. The plant configuration that is
required to perform this test (refueling) would clearly place the
Haddam Neck Plant in a state that would be able to accept all possible
ABT scheme test outcomes, normal and abnormal. Therefore, these
proposed changes do not result in any reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Russell Library, 123 Broad
Street, Middletown, Connecticut 06457.
Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &
Howard, Counselors at Law, City Place, Hartford, Connecticut 06103-
3499.
NRC Project Director: John F. Stolz.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of amendment request: October 29, 1993, as supplemented on
March 28, 1994.
Description of amendment request: This amendment is an additional
followup to the amendment request of May 29, 1992, published in the
Federal Register (57 FR 30242) on July 8, 1992, which changed the
Technical Specifications Section 1.0, Definitions, to accommodate a 24-
month fuel cycle and which proposed the extension of the test intervals
for specific surveillance tests. This amendment proposes extending the
surveillance intervals to 24 months for the following additional
surveillance tests:
(1) Volume Control Tank Level Transmitter
(2) Containment High Range Area Radiation Monitors, R-25 and R-26
(3) Safety Injection System Electrical Loading
(4) Safety Injection (SI) System
(5) Reactor Coolant System Sub-Cooling Margin Monitor
The changes requested by the licensee are in accordance with
Generic Letter 91-04, ``Changes in Technical Specification Surveillance
Intervals to Accommodate a 24-Month Fuel Cycle.''
The October 29, 1993, submittal included surveillance tests for the
Auxiliary Feedwater System which duplicated a previous request which
was subsequently approved. The March 28, 1994, submittal withdrew the
duplicated request.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazard
consideration, which is presented below:
(1) Volume Control Tank Transmitter:
The proposed change does not involve a significant hazard
consideration since:
1. A significant increase in the probability or consequences of an
accident previously evaluated will not occur.
It is proposed that the channel calibration frequency for the
volume control tank level instrumentation be changed from every 18
months (+25%) to every 24 months (+25%).
A statistical analysis of channel uncertainty for a 30 month
operating cycle has been performed. Based upon this analysis it has
been concluded that sufficient margin exists between the existing
Technical Specification limits and the licensing basis Safety Analysis
limits to accommodate the channel statistical error resulting from a 30
month operating cycle. The existing margin between the Technical
Specification limit and the Safety Analysis limit provides assurance
that plant protective actions will occur as required. It is therefore
concluded that changing the surveillance interval from 18 months (+25%)
to 24 months (+25%) will not result in a significant increase in the
probability or consequences of an accident previously evaluated.
2. The possibility of a new or different kind of accident from any
accident previously evaluated has not been created.
The proposed change is operating cycle length from a maximum of
22.5 months to 30 months resulting from an increased surveillance
interval will not result in a channel statistical allowance which
exceeds the current margin between the existing Technical Specification
limits and the Safety Analysis limits. Plant equipment, which will be
set at (or more conservatively than) Technical Specification limits,
will therefore provide protective functions to assure that Safety
Analysis limits are not exceeded. This will prevent the possibility of
any new or different kind of accident from that previously evaluated
from occurring.
3. A significant reduction in a margin of safety is not involved.
The change in surveillance interval from a maximum of 22.5 months
to 30 months resulting from an increased operating cycle will not
result in a channel statistical allowance which exceeds the margin
which exists between the current Technical Specification limit and the
licensing basis Safety Analysis limit. This margin, which is equivalent
to the existing margin, is necessary to assure that protective safety
functions will occur so that Safety Analysis limits are not exceeded.
(2) Containment High Range Area Radiation Monitors, R-25 and R-26:
The proposed change does not involve a significant hazard
consideration since:
1. There is no significant increase in the probability or
consequences of an accident.
It is proposed that the calibration frequency for the high-range
Containment Radiation Monitors (R-25 and R-26) be revised from every 18
months (+25%) to every 24 months (+25%).
These two monitors are redundant to each other and are used for
post accident monitoring purposes. They serve no function during normal
plant operation. Furthermore, they serve no purpose in preventing
accident initiation or mitigation. They are used for Emergency Planning
purposes to indicate a release of radioactivity to containment.
Review of past test results indicates that the devices have proven
reliable during past surveillances and there was no indication that
they would not remain operable for an extended operating cycle. In
addition, the devices are essentially redundant to each other. Each
device would respond to a release of radioactivity to Containment.
In consideration that the monitors are redundant, and in view of
the past test history of the monitors, there would be no significant
increase in the probability or consequences of an accident due to an
extended operating cycle.
2. The possibility of a new or different kind of accident from any
previously analyzed has not been created.
The role of R-25 and R-26 is in the assessment of radiological
releases to Containment. In this function it is important that one of
the instruments, being high range, respond to a radiological release.
Indications from the devices are not used in a quantitative manner.
Rather they are used for qualitative purposes. Due to redundancy and
past test history, continued operability is expected. In addition, the
instruments serve no function in preventing accident initiation or
accident mitigation. Therefore, it is concluded that an extended
operating cycle for these monitors would not result in the possibility
of a new or different kind of accident from any previously analyzed.
3. There has been no significant reduction in the margin of safety.
Due to the qualitative function served by these two instruments as
well as their redundancy and acceptable past test history, no
significant reduction in the margin of safety due to an extended
operating cycle is expected.
(3) Safety Injection System Electrical Loading
The proposed change does not involve a significant hazard
consideration since:
1. There is no significant increase in the probability or
consequences of an accident.
The test procedure under consideration is one of the more
complicated surveillance procedures accomplished at a refueling
interval. Considering the vast number of components that are tested it
is highly improbable that some deficiencies will not occur. When such
problems are encountered it is important to note whether the failure is
time dependent and, in addition, whether the corrective maintenance
implemented prevents recurrences in the future. In consideration of the
evaluation of past test observations it is important to note that the
problems which occurred were not time dependent and that maintenance
practices have been effective in precluding future failures of the same
type. Equally important is whether the emergency power system would
have performed its intended safety function if the situation was not a
test but represented an actual demand upon the system. Test acceptance
criteria are always more stringent than required by accident scenarios
to provide margin. As discussed above the two most significant findings
were a failure of a CCW [Component Cooling Water] pump to strip from
the bus during the 1989 test and a relay which did not function within
its timing sequence. In the first instance, the diesel generator was
not overloaded. In the second instance, the relay did function albeit
not within the allotted time. In both cases, safety functions would
have been performed.
Thus, it is concluded that an extended period between surveillances
will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The possibility of a new or different kind of accident from any
previously analyzed has not been created.
The deficiencies noted in the test data taken during the last
several refueling outage surveillances were not substantial in number
and would not have impacted the capability of the Safety Injection
System and its emergency power supply to perform its intended safety
function. The effectiveness of maintenance practices, both preventive
and corrective, has been proven in that deficiencies noted in one test
are not repeated in subsequent tests. The last refueling surveillance
test was completely successful where no new test failures were noted.
Because past test deficiencies do not appear to be time dependent,
extending the surveillance interval by 7.5 months is not expected to
create the possibility of a new or different kind of accident from any
accident previously created.
3. There has been no reduction in the margin of safety.
Because previous tests indicate that the engineered safety features
power supply would have performed its safety function if called upon
over the past several years, it is concluded that extending the
operating cycle by several months will not involve a significant
reduction in a margin of safety.
(4) Safety Injection System
The proposed change does not involve a significant hazard
consideration since:
1. There is no significant increase in the probability or
consequences of an accident.
The central safety objective in reactor design and operation is the
control of reactor fission products from the fuel. Four methods are
used to ensure this objective. Two of these methods are: (1) Retention
of fission products in the reactor coolant for whatever leakage occurs;
and (2) retention of fission products by the containment for
operational and accidental releases beyond the reactor coolant
boundary. The engineered safety features are the provisions in the
plant that embody these two methods to prevent the occurrence or to
ameliorate the effects of serious accidents.
The engineered safety features systems are the containment system,
the safety injection system, the containment spray system, the
containment air recirculation cooling and filtration system, the
isolation valve seal-water system, and the containment penetration and
weld channel pressurization system. Each engineered safety feature
provides sufficient performance capability to accommodate any single
failure of an active component and still function in a manner to avoid
undue risk to the health and safety of the public.
A comprehensive program of plant testing is formulated for all
equipment, systems, and system control vital to the functioning of
engineered safety features. The program consists, in part, of
integrated tests of the systems as a whole and periodic tests of the
actuation circuitry and mechanical components.
An assessment has been performed of the test results from the last
five refueling outages, covering a period in excess of seven years. In
reviewing the test results particular attention was directed towards
those test anomalies which directly impacted test acceptance criteria
and, thus, influence the capability of the safety injection system to
perform its intended safety function. Although in each test a problem
area was identified, the number of such events were minimal.
Furthermore, after corrective action these events were not repeated in
subsequent system tests. In all instances the problems were not
identified to be time dependent. Furthermore, the consequence from a
system safety function perspective was minimal. Thus, it is concluded
that extending the surveillance interval by several months will not
involve a significant increase in the probability or consequences of an
accident previously evaluated.
2. The possibility of a new or different kind of accident from any
previously analyzed has not been created.
The number of problem areas in each test have been few and of
minimal to nonexistent safety significance. In 1986, a valve failure
occurred which would have been detected by alternate means during an
extended operating cycle. In another instance, lack of valve movement
could not be repeated in a second test, leading to the conclusion that
the valve malfunction was not induced by the system but was the result
of the test process. In the last problem area, manual SI initiation, no
credit is taken within the FSAR [Final Safety Analysis Report] accident
analysis for this function. In 1989, a series of containment isolation
valves failed to stroke as required. In three instances the valves
failed closed, which is the correct position. In the other instances,
either the redundant valve did stroke to the correct position or the
valve was located in a closed system. In all these events there was
minimal impact upon safety. More importantly, after corrective action,
these failures were not repeated in the 1991 or 1993 tests. In 1991,
one breaker failed to perform within specifications and thus was
considered defective. In 1993 there were no major equipment
malfunctions, although one containment isolation valve failed to
perform as required.
In summary, although there have been anomalies in all of the tests
evaluated, none were deemed serious enough to impact the safety
function of the safety injection system or to be considered as having a
negative affect upon an increased interval of several months between
surveillances. Therefore, it has been concluded that an increased
operating cycle will not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. There has been no reduction in the margin of safety.
The results of the previous five cycles of test data have been
evaluated. None of the anomalies observed were sufficiently serious to
impact the performance of the Safety Injection System or to weigh
against an extended operating cycle. As there are no other changes to
the safety analysis parameters which are impacted by an extended
interval between surveillances, it is concluded that this change will
not involve a significant reduction in the margin of safety.
(5) Reactor Coolant System Sub-Cooling Margin Monitor:
The proposed change does not involve a significant hazards
consideration since:
1. There is no significant increase in the possibility or
consequences of an accident.
It is proposed that the channel calibration frequency for the
volume control tank instrumentation be changed from every 18 months
(+25%) to every 24 months (+25%).
The sub-cooling margin monitoring function is not relied upon
during normal operation. There is no reference to its use in the Indian
Point Unit 2 standard operating procedures. No credit is taken for this
monitoring function within the safety analysis for either the
prevention or mitigation of an accident. The increase in ``normal''
operating uncertainty, due to the longer operating cycle, as well as
``adverse'' uncertainties, is being incorporated in the EOPs [Emergency
Operating Procedures]. Therefore, the slight increase in uncertainty
associated with a longer operating cycle between surveillances will not
cause a significant increase in the probability or consequences of an
accident.
2. The possibility of a new or different kind of accident from any
previously analyzed has not been created.
The sub-cooling margin serves no purpose during normal operation
for prevention of an accident. No credit is taken within the FSAR
Safety Analysis for accident mitigation. The sub-cooling margin monitor
is relied upon within the Emergency Operating Procedures. Thus, the
normal uncertainty due to a 30 month operating cycle, as supplemented
by the instrument loop error due to a post-accident harsh environment,
is being factored into the Emergency Operation Procedures in accordance
with Emergency Response Guidelines. Thus, it is concluded that the
possibility of a new or different kind of accident from any previously
analyzed has not been created.
3. There has been no reduction in the margin of safety.
Because the sub-cooling margin monitor serves no purpose during
normal operation and appropriate measures have been implemented to
reflect the additional uncertainty due to a 30 month operating cycle
into the EOPs, it is concluded that there will be no significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,
New York, New York 10003.
NRC Project Director: Michael L. Boyle, Acting.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of amendment request: June 1, 1994.
Description of amendment request: The proposed amendment would
revise the Technical Specifications to allow extended Rod Position
Indication (RPI) deviation limits and on-line calibration of the RPI
channels. Specifically, Section 3.10.6.1 would be changed to allow
extended RPI deviation limits, and Section 3.10.4.4 would be changed to
allow on-line calibration of the RPI channels. The Basis for Section 10
would be changed to reflect the above, and in addition, Section
3.10.6.2 would be changed to clarify the operability requirements
during calibration. The proposed changes to Sections 3.10.6.1 and
3.10.4.4 include power limits to be included in the Core Operating
Limit Report (COLR). The use of a COLR for cycle specific core
operating limits was proposed by the licensee by submittal of October
29, 1993, and is currently under review by the NRC staff.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident previously
evaluated?
Response: Neither the probability nor the consequences of an
accident previously analyzed is increased due to the proposed changes.
All peaking factors will remain within the limits of the Technical
Specifications. Both the shutdown margin and the axial flux difference
will be maintained within the limits of the Technical Specifications.
There will be no fuel damage due to the changes. All design and safety
criteria will be met.
2. Does the proposed license amendment create the possibility of a
new or different kind of accident from any previously evaluated?
Response: The changes will not create the possibility of a new or
different kind of accident. The calibration will be performed using
plant procedures that have been reviewed and approved by Con Edison's
Safety Committees. It has been shown that even with the new RPI
deviation bands and on-line calibration, all power distribution limits
will be met.
3. Does the proposed amendment involve a significant reduction in
the margin of safety?
Response: The proposed amendment does not involve a significant
reduction in the margin of safety. There will be no change in the power
distribution limits used in the design and safety analyses and the
required shutdown margin will be maintained. It has been shown that
there is no fuel failure as a result of this change.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,
New York, New York 10003.
NRC Project Director: Michael L. Boyle.
Detroit Edison Company, Docket No. 50-16, Enrico Fermi Power Plant,
Unit 1, Monroe County, Michigan
Date of application for amendment: December 9, 1993 (Reference NRC-
93-0143)
Brief description of amendment: This Licensee Amendment Request
(LAR) proposes to revise the Enrico Fermi Power Plant, Unit 1,
Technical Specifications (TS) to bring the TS into conformance with a
revision of 10 CFR Part 20 (56 FR 23360).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
a. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes are administrative in nature and do not impact
the SAFSTOR status or design of any plant structures, systems or
components. As a result, this proposed change cannot increase the
probability or the consequences of any accident previously evaluated.
b. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes do not affect the plant SAFSTOR status as
defined. As a result, the proposed changes cannot create the
possibility of a new or different kind of accident from any accident
previously evaluated.
c. Does the change involve a significant reduction in a margin of
safety?
The proposed changes do not affect the plant SAFSTOR status. The
changes will not increase the amounts or change the types of effluents
that may be released offsite. These changes only ensure compliance with
revised 10 CFR 20. These changes do not alter any of the requirements
or responsibilities for protection of the public against radiation
hazards. As a result, these changes do not reduce the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161.
Attorney for licensee: John Flynn, Esq., Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan 48226.
NRC Branch Chief: John H. Austin.
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver
Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania
Date of amendment request: September 16, 1992, as superseded
February 4, 1994.
Description of amendment request: The proposed amendment would
supersede in its entirety a previous proposed amendment which was
submitted by letter dated September 16, 1992. A notice of application
and proposed no significant hazards consideration determination for the
September 16, 1992, submittal was published in the Federal Register on
January 21, 1993 (58 FR 5429); this notice supersedes the January 21,
1993, notice in its entirety.
The proposed amendment would modify the Technical Specifications
(TSs) related to containment air locks to make them as close to the
Improved Standard TSs in NUREG-1431 as the plant-specific design will
permit. The proposed changes in TS 3.6.1.1 and 3.6.1.3 would modify
surveillance requirements and limiting conditions of operation and
effect numerous administrative and format changes. The changes relate
to air lock operability, leak testing, and door interlocks.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The probability of occurrence of a previously evaluated accident is
not increased because the containment air locks do not effect the
initiation of any design basis accident [DBA]. The consequences of an
accident are also not significantly increased because the proposed
revisions to the action statements will continue to ensure that at
least one door in each air lock is maintained closed. A single door in
each air lock is capable of withstanding a pressure in excess of the
maximum expected pressure following a DBA. The structural integrity and
leak tightness of the containment will not be changed by this proposed
revision. For the brief period of time that the operable air lock door
is open and the inoperable door is providing the single containment
barrier, the consequences of [2an] accident may be increased. However,
the probability of an event occurring requiring containment integrity
is sufficiently remote to justify limited access when required.
Therefore, based on the continued ability of the containment air
locks to provide a barrier to limit leakage from containment during a
DBA, this proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Air lock operation does not interface with the reactor coolant
pressure boundary or any other mechanical or electrical controls which
could impact the operations of the reactor or its direct support
systems.
Containment air locks are designed for the purpose of containment
entry and exit. During this operation, the air lock maintains
containment integrity by providing at least one door which is capable
of providing a leak tight barrier during a DBA.
The proposed changes will continue to ensure that air lock
operation is performed as assumed in the original design of the plant.
During the period when the operable door is open and the other door
inoperable, at least one door is being maintained closed as designed.
This condition is ensured due to the subatmospheric conditions that
exist during plant operation. The operable air lock door cannot be
safely opened unless the inoperable door is closed due to the
approximately 5 psi pressure differential that exists. The operable air
lock door would only be opened long enough to allow personnel to enter
the air lock.
Therefore, this proposed change does not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin of
safety?
The applicable margin of safety consists of maintaining the primary
containment leak rates within the assumptions of the DBA analysis.
These leak rates are maintained provided at least one operable air lock
door remains closed during the event.
The proposed revisions will continue to ensure that at least one
air lock door is maintained closed. During the brief period of time
that an operable air lock door is open and the inoperable door is
providing the single containment barrier, the margin of safety is
decreased. The inoperable door may not limit containment leak rates
within the assumptions of the DBA analysis. However, the probability of
an event requiring the inoperable air lock door to limit containment
leakage occurring during this time period is sufficiently low and the
overall margin of safety would not be decreased by a significant
amount. The proposed increase in allowable door seal leakage will not
affect the overall ability of the containment air locks to restrict the
release of fission products to the environment. The overall air lock
leakage limit of less than or equal to .05 La remain unchanged.
The amount of leakage which the air lock(s) are permitted to contribute
to the combined containment leakage limit of 0.60 La remain
unchanged. Therefore, the margin of safety due to increasing the door
seal leakage limit remains unchanged.
Therefore, this proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: B.F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.
Attorney for licensee: Gerald Charnoff, Esquire, Jay E. Silberg,
Esquire, Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW.,
Washington, DC 20037.
NRC Project Director: Walter R. Butler.
Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric
Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-424 and
50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County,
Georgia
Date of amendment request: March 31, 1994.
Description of amendment request: The proposed amendments would
change Technical Specification 3/4.7.1.1 and its Bases regarding
maximum allowed reactor thermal power operation with inoperable main
steam safety valves.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The new power range neutron flux high setpoint values will ensure that
the secondary side steam pressure will remain below 110 percent of its
design value following a loss of load/turbine trip (LOL/TT) when one or
more main steam safety valves (MSSVs) are declared inoperable.
Therefore, this transient will remain classified as a Condition II
probability event (faults of moderate frequency) per ANSI--N18.2, 1973
as discussed in Section 15.0.1 of the VEGP Final Safety Analysis Report
(FSAR). Accordingly, since the new power range setpoints will maintain
the capability of the MSSVs to perform their pressure relief function
associated with a LOL/TT event, there will be no effect on the
probability or consequences of an accident previously evaluated.
2. The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated. The
proposed changes do not involve any change to the configuration or
method of operation of any plant equipment, and no new failure modes
have been defined for any plant system or component. The new power
range neutron flux high setpoints will maintain the capability of the
MSSVs to perform their pressure relief function to ensure the secondary
side steam design pressure is not exceeded following a LOL/TT.
Therefore, since the function of the MSSVs is unaffected by the
proposed changes, the possibility of a new or different kind of
accident from any accident previously evaluated is not created.
3. The proposed changes do not involve a significant reduction in a
margin of safety. The algorithm methodology used to calculate the new
power range neutron flux high setpoints is conservative and bounding
since it is based on a number of inoperable MSSVs per loop; i.e., if
only one MSSV in one loop is out of service, the applicable power range
setpoint would be the same as if one MSSV in each loop were out of
service. Another conservatism with the algorithm methodology is with
the assumed minimum total steam flow rate capability of the operable
MSSVs. The assumption is that if one or more MSSVs are inoperable per
loop, the inoperable MSSVs are the largest capacity MSSVs, regardless
of which capacity MSSVs are actually inoperable. Therefore, since the
power range setpoints calculated for the proposed changes using the
algorithm methodology are more conservative and ensure the secondary
side steam design pressure is not exceeded following a LOL/TT, there
will not be a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Burke County Public Library,
412 Fourth Street, Waynesboro, Georgia 30830.
NRC Project Director: David B. Matthews.
Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric
Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-424 and
50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County,
Georgia
Date of amendment request: May 20, 1994.
Description of amendment request: The proposed amendments would
relocate the heat flux hot channel factor penalty of 2 percent in
Specification 4.2.2.2.f to the cycle-specific Core Operating Limits
Report to allow burnup-dependent values of the penalty in excess of 2
percent. The licensee also proposes to revise the reference in
Specification 6.8.1.6 to the Westinghouse FQ(Z) surveillance
methodology in order to reflect Revision 1 of WCAP-10216-P,
``Relaxation of Constant Axial Offset Control--FQ Surveillance
Technical Specification,'' approved by the NRC on November 26, 1993.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change involves only the manner in which the penalty
factors for FQ(Z) would be specified (i. e., a burnup-dependent
factor specified in the Core Operating Limits Report (COLR) versus a
constant factor specified in the TS [Technical Specification]). This is
simply used to account for the fact that FQ(Z) may increase
between surveillance intervals. These penalty factors are not assumed
in any of the initiating events for the accident analyses. Therefore,
the proposed change will have no effect on the probability of any
accidents previously evaluated. The penalty factors specified in the
COLR will be calculated using NRC-approved methodology and will
therefore continue to provide an equivalent level of protection as the
existing TS requirement. Therefore, the proposed change will not affect
the consequences of any accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not involve a physical alteration to the
plant (no new or different kind of equipment will be installed) or
alter the manner in which the plant would be operated. Thus, this
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does this change involve a significant reduction in a margin of
safety?
The proposed change will continue to ensure that potential
increases in FQ(Z) over a surveillance interval will be properly
accounted for. The penalty factors will be calculated using NRC-
approved methodology. Therefore the proposed change will not involve a
reduction in margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Burke County Public Library,
412 Fourth Street, Waynesboro, Georgia 30830.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308.
NRC Project Director: David B. Matthews.
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: June 22, 1994.
Description of amendment request: The proposed amendment changes
Technical Specification Sections 1.6, 3.2.A, 3.9.F.5, and 4.2.A which
specify the Shutdown Margin (SDM) requirements that ensure the reactor
can be made subcritical and can be maintained sufficiently subcritical
to preclude inadvertent criticality in any core condition. The
amendment also proposes a new definition, Shutdown Margin, Section
1.45. The proposed changes address the requirements for SDM
demonstration and provide clarification for actions if SDM is not met.
The amendment also proposes administrative changes to Sections 1.7
and 3.2.B.2 (b). The definition, COLD SHUTDOWN CONDITION, was
simplified by stating the reactor is in the SHUTDOWN CONDITION which
eliminates the need of repeating the requirements for this condition.
The note which permitted unlimited reactor startups without the Rod
Worth Minimizer during Cycle 11 is no longer applicable. The note and
its reference are deleted from the new page 3.2-2. Starting with page
3.2-2 in Section 3.2, the pages were renumbered and repaginated to
accommodate the changes in text.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
GPU Nuclear has determined that operation of the Oyster Creek
Nuclear Generating Station in accordance with the proposed Technical
Specifications does not involve a significant hazard. The changes do
not:
1. Involve a significant increase in the probability or the
consequence of an accident previously evaluated.
The proposed SDM Limits are more restrictive and provide adequate
shutdown margin for various modes of reactor operation. Since the new
SDM limits do not modify any initial conditions for the accidents
previously evaluated in the SAR [Safety Analysis Report], the proposed
changes do not involve a significant increase in the probability or
consequences of these accidents.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
The proposed TS changes do not modify the function of any
structure, system or component. The new Shutdown Margin requirements
will still meet the basic criterion that the core in its maximum
reactivity condition be subcritical with the control rod of highest
worth fully withdrawn and all operable rods fully inserted. Based on
these facts, the proposed TS changes do not create the possibility of a
new or different kind of accident from any previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes do not reduce the margin of safety, because
the new SDM limits where the highest worth control rod is determined
analytically (0.38% delta k) or by measurement (0.28% delta k) are more
restrictive than the current Oyster Creek limit (0.25% delta k).
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, New Jersey
08753.
Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz.
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas, Docket
Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda
County, Texas
Date of amendment request: June 6, 1994.
Description of amendment request: The licensee proposes to modify
the South Texas Project, Units 1 and 2, Technical Specification 3/
4.8.1.1, ``A.C. Sources,'' to revise the action statements and
surveillance requirements for testing of the standby diesel generators
(SDGs). The proposed amendment would eliminate excessive and
unnecessary testing of the SDGs consistent with the guidance provided
in NUREG-1366, ``Improvements to Technical Specifications Surveillance
Requirements,'' NUREG-1431, ``Standard Technical Specifications for
Westinghouse Plants,'' Generic Letter 84-15, ``Proposed Staff Actions
to Improve and Maintain Diesel Generator Reliability,'' and Generic
Letter 93-05, ``Line-Item Technical Specifications Improvements to
Reduce Surveillance Requirements for Testing During Power Operation.''
This request replaces a request for amendment dated November 23, 1993,
which was noticed on January 5, 1994 (59 FR 621). This revised
amendment request includes elimination of additional identified
unnecessary testing discovered since the original submittal. The
changes include: (1) eliminating the requirement to demonstrate the
operability of an operable SDG whenever an offsite AC power source is
determined to be inoperable, or whenever a support system or an
independently testable component of another SDG is inoperable, (2)
eliminating the requirement to load the diesel in 10 minutes during
testing, (3) replacing the minimum required loading for testing with a
load band, (4) relocating some surveillance requirements to the Diesel
Fuel Oil Testing Program, and (5) eliminating unnecessary loss of
offsite power tests.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The Standby Diesel Generators do not initiate any accidents,
therefore these changes do not increase the probability or [of] an
accident previously evaluated. The proposed changes will permit the
elimination of the unnecessary mechanical stress and wear on the diesel
engine and generator while ensuring that the diesel generators will
perform their designed function. The elimination of this mechanical
stress and wear will improve the reliability and availability of the
Standby Diesel Generators which will have a positive effect on the
ability of the diesel generators to perform their design function.
Therefore, the consequences of an accident previously evaluated are not
increased. The proposed changes are consistent with NUREG-1366, NUREG-
1431, Generic Letter 93-05, and Generic Letter 84-15.
2. The proposed change does not create the possibility of a new or
different kind of accident from any previously evaluated.
The elimination of these unnecessary tests does not affect the
design bases of the SDGs, or any of the accident evaluations involving
the SDGs. The SDGs are designed to provide electrical power to the
equipment important for safety during all modes and plant conditions
following a loss of offsite power. The test schedule established in
accordance with GL 84-15 assures that operable SDGs are capable of
performing their intended safety function. The proposed changes to the
surveillance requirements are consistent with NUREG-1431, NUREG-1366,
Generic Letter 93-05, industry operating experience, and South Texas
Project operating experience. These changes are intended to improve
plant safety, decrease equipment degradation, and remove unnecessary
burden on personnel resources by reducing the amount of testing that
the Technical Specification requires during power operation. Relocating
the diesel fuel oil testing requirement to the STP Fuel Oil Monitoring
Program outside of the Technical Specifications is an administrative
change consistent with NUREG-1431 and consequently has no effect on
accident probability, consequences, or margin. Therefore, this change
does not create the possibility of a new or different kind of accident
from any previously evaluated.
3. The proposed change does not involve a significant reduction in
a margin of safety.
The proposed changes extend testing frequency and eliminate
unnecessary mechanical stress and wear on the diesel generator in an
effort to improve plant reliability and safety. These changes are
consistent with NUREG-1431, NUREG-1366, industry operating experience,
and STP operating experience and do not adversely affect the design
bases, accident analysis, reliability or capability of the SDGs to
perform their intended safety function. Relocating the diesel fuel oil
testing requirements to the STP Fuel Oil Monitoring Program outside of
the Technical Specifications is an administrative change consistent
with NUREG-1431 and consequently has no effect on accident probability,
consequences, or margin. Therefore the proposed changes do not involve
any reduction in a margin to safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involve no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas
77488.
Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger,
P.C., 1615 L Street, N.W., Washington, D.C. 20036.
NRC Project Director: William D. Beckner.
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn
County, Iowa
Date of amendment request: June 18, 1993 as supplemented on
December 17, 1993 and May 5, 1994.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) by clarifying TS wording for
the Low Pressure Coolant Injection (LPCI) and Containment Spray modes
of the Residual Heat Removal (RHR) system to assure consistency with
requirements of the DAEC Updated Final Safety Analysis Report.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) The probability or consequences of a previously-analyzed
accident will not be increased by these proposed changes to the LPCI
and Containment Spray LCOs and BASES because they merely clarify
existing TS requirements and are consistent with the DAEC UFSAR
accident analysis. The addition of the footnote clarifying LPCI
OPERABILITY during RHR system operation in the Shutdown Cooling mode is
consistent with the requirements in the NRC Standard TS (NUREG-1433).
No changes in either system design or operating strategies will be made
as a result of these changes, thus no opportunity exists to increase
the probability or consequences of a previously-analyzed accident.
(2) The possibility of a new or different kind of accident from
those previously analyzed will not be created by these changes to the
LPCI and Containment Spray LCOs and BASES because they merely clarify
existing requirements. The addition of the footnote clarifying LPCI
OPERABILITY during RHR system operation in the Shutdown Cooling mode is
consistent with the requirements in the NRC Standard TS (NUREG-1433).
No changes in either system design or operating strategies will be made
as a result of these changes, thus no possibility exists to introduce a
new or different kind of accident.
(3) The margin of safety will not be decreased as a result of these
changes because they merely clarify existing TS requirements and are
consistent with the UFSAR accident analysis. The addition of the
footnote clarifying LPCI OPERABILITY during RHR system operation in the
Shutdown Cooling mode is consistent with the requirements in the NRC
Standard TS (NUREG-1433). No changes in either system design or
operating strategies will be made as a result of these changes, thus no
possibility exists to reduce a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, S.E., Cedar Rapids, Iowa 52401.
Attorney for licensee: Jack Newman, Esquire, Kathleen H. Shea,
Esquire, Newman and Holtzinger, 1615 L Street NW., Washington, DC
20036.
NRC Project Director: John N. Hannon.
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point
Nuclear Station, Unit 2, Oswego County, New York
Date of amendment request: July 1, 1994.
Description of amendment request: The proposed amendment would
revise the drawdown time testing requirement of Technical Specification
(TS) 4.6.5.1.c.1 and the secondary containment inleakage testing
requirement of TS 4.6.5.1.c.2. These revisions would support a revised
design basis radiological analysis which would support an increase in
secondary containment drawdown time from 6 to 60 minutes by taking
credit for fission product scrubbing and retention in the suppression
pool. The current design basis radiological analysis does not take
credit for the pressure suppression pool as a fission product cleanup
system as permitted in NUREG-0800, Section 6.5.5, ``Pressure
Suppression Pool as a Fission Product Cleanup System.'' The proposed
amendment would also take credit for additional mixing of primary
containment and engineered safety feature systems leakage with 50
percent of the secondary containment free air volume prior to the
release of radioactivity to the environment. In the revised analysis,
mixing is assumed to occur at the onset of a Design Basis Loss-of-
Coolant Accident as the primary containment and the engineered safety
feature systems leak into secondary containment. The current analysis
takes credit for mixing within secondary containment only after
achieving a -0.25 inch water gauge (WG) pressure in secondary
containment with respect to the outside surrounding atmosphere. The
licensee's radiological evaluation for this accident, which reflects
these proposed changes and an assumed drawdown time of 60 minutes, has
determined that the radiological doses remain below 10 CFR Part 100
guidelines values and General Design Criterion 19 criteria. The revised
radiological doses, as calculated by the licensee, are lower than the
doses currently presented in the Updated Safety Analysis Report (USAR).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The operation of the Nine Mile Point Unit 2, in accordance with the
proposed amendment, will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Secondary containment and the SGTS [Standby Gas Treatment System]
are not initiators or precursors to an accident. Secondary containment
provides a pressure boundary, with limited inleakage, for the purpose
of establishing a negative pressure to prevent a ground level
unfiltered release of radioactivity. SGTS responds to accidents
involving a release of radioactivity to the secondary containment by
establishing and maintaining a negative pressure inside secondary
containment and by providing an elevated filtered release. Therefore,
changes to SECONDARY CONTAINMENT INTEGRITY surveillances cannot affect
the probability of a previously evaluated accident.
The suppression pool and secondary containment are largely passive
in nature, and the active components are suitably redundant. Therefore,
their fission product attenuation functions can be accomplished
assuming a single failure.
Currently, using an assumed drawdown time of 6 minutes, the
radiological doses for a DBA-LOCA [Design Bases Accident--Loss-of-
Coolant Accident] are below the guidelines of 10 CFR Part 100 and GDC
[General Design Criterion] 19 criteria. The calculated doses,
considering the pressure suppression pool as a fission product cleanup
system, additional mixing within secondary containment and an assumed
secondary containment drawdown time of 60 minutes, are lower than the
previously calculated doses. The new doses are below 10 CFR Part 100
guideline values and GDC 19 criteria. The revised radiological analysis
follows the source term assumptions of RG [Regulatory Guide] 1.3, with
the exception of regulatory position C.1.f as permitted by SRP
[Standard Review Plan] Section 6.5.5, and continues to provide a
conservative representation of the timing and the composition of the
release of radioactivity from secondary containment during a DBA-LOCA.
The Technical Specification SRs [Surveillance Requirements] will
ensure a continued state of readiness for the SGTS, the secondary
containment, the suppression pool and the suppression chamber/drywell
vacuum breakers. Therefore, the assumptions used in the dose assessment
will continue to bound the actual bypass of the suppression pool and
the mixing in secondary containment during a DBA-LOCA. The proposed
changes to the surveillances provide assurance that the performance of
the SGTS and secondary containment supports the radiological analysis.
Accordingly, as shown in Table 1, page 14 of 20, [of the July 1, 1994,
amendment request] operation with the SGTS and the proposed change to
the surveillances for secondary containment will not significantly
increase the consequences of an accident previously evaluated.
The operation of Nine Mile Point Unit 2, in accordance with the
proposed amendment, will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed change to the surveillances ensures that the SGTS and
secondary containment will be available to respond to an accident such
that the guidelines of 10 CFR Part 100 and the limits of GDC 19 are not
exceeded. The proposed change to the surveillances reflect
consideration of the pressure suppression pool as a fission product
cleanup system and credit for additional mixing in secondary
containment. The suppression pool will continue to perform its safety
functions as a pressure suppression pool and as a source of water to
support emergency core cooling system operation during a DBA-LOCA. In
addition, secondary containment will continue to perform its safety
function of controlling and minimizing radioactive leakage to the
outside atmosphere during a DBA-LOCA. Safety related equipment will
continue to be OPERABLE in the radioactive environment of secondary
containment to mitigate the consequences of a DBA-LOCA. Accordingly,
the proposed Technical Specification change will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The operation of Nine Mile Point Unit 2, in accordance with
proposed amendment, will not involve a significant reduction in a
margin of safety.
The SGTS exhausts the secondary containment atmosphere to the
environment through the filtration system. To verify the SGTS has not
degraded, SR 4.6.5.1.c.1 verifies that each SGTS subsystem will
establish and maintain a pressure in the secondary containment that is
equal to or more negative than -0.25 inch WG within the required time
limit. To verify secondary containment is intact, SR 4.6.5.1.c.2
demonstrates that one SGTS subsystem can maintain a pressure which is
equal to or more negative than -0.25 inch WG for 1 hour at a flow rate
less than or equal to the maximum allowed inleakage. The 1 hour test
period allows secondary containment to be in thermal equilibrium at
steady state conditions. Furthermore, as an interim measure, NMPC
[Niagara Mohawk Power Corporation] implemented certain compensatory
measures through administrative controls to ensure that the
radiological consequences of a DBA-LOCA would remain within regulatory
criteria. Together, these tests and the compensatory measures assure
SGTS performance and secondary containment boundary integrity.
The proposed change to these surveillances incorporate changes to
the design basis, i.e., credit for fission product scrubbing and
retention by the suppression pool and credit for additional mixing
within secondary containment. The new inleakage is 2670 cfm which is
loss [less] than one change of the secondary containment free air
volume per day. The new drawdown time limit reflects consideration of
the proposed change in the secondary containment inleakage limit. Due
to the effects of service water temperature, inside and outside
temperature, flow measurement inaccuracies and actual test pressures,
meeting the current SRs does not by itself assure adequate SGTS
performance. Therefore, the surveillances' results are adjusted to
account for actual test conditions. Compliance with the proposed
surveillances assures that the SGTS can achieve and maintain -0.25 inch
WG in less than 60 minutes following a postulated DBA-LOCA. Achieving
-0.25 inch WG within 60 minutes assures that radiological doses will
remain below regulatory limits (see Table 1 [of the July 1, 1994,
amendment request]). Therefore, the proposed surveillances, together
with the proposed adjustments, provide adequate assurance of SGTS
performance and secondary containment boundary integrity. Accordingly,
the proposed Technical Specification change will not involve a
significant reduction in a margin of safety.
Therefore, as determined by the analysis above, this proposed
amendment involves no significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street NW., Washington, DC 20005-3502.
NRC Project Director: Michael L. Boyle.
Northern States Power Company, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of amendment request: June 8, 1994.
Description of amendment request: The proposed amendment would
revise sections 3.7/4.7, which pertain to the Standby Gas Treatment
System (SGTS) and Secondary Containment. The proposed amendment would
revise the surveillance requirements for both SGTS and secondary
containment and revise the performance requirements for the SGTS
filters and process stream electric heaters.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment will not involve a significant increase
in the probability or consequences of an accident previously evaluated.
The function of the SGTS and secondary containment is to mitigate
the consequences of a loss of coolant accident and fuel handling
accidents. The proposed changes maintain or improve this capability.
Therefore, this amendment will not cause a significant increase in the
probability or consequences of an accident previously evaluated for the
Monticello plant.
2. The proposed amendment will not create the possibility of a new
or different kind of accident from any accident previously analyzed.
The proposed changes to Technical Specifications for the standby
gas treatment system and secondary containment do not alter the
function of the systems or its interrelationships with other systems.
The proposed changes provide requirements to ensure the systems are
capable of performing the required functions or that actions are taken
to minimize the potential for its function being required consistent
with regulatory guidance; therefore, this amendment will not create the
possibility of a new or different kind of accident from any accident
previously analyzed.
3. The proposed amendment will not involve a significant reduction
in the margin of safety.
Improvements in the margin of safety are provided via the permanent
elimination of a potential single failure which could adversely affect
both standby gas treatment systems by deleting the reference to the
standby gas system room heaters in the technical specification bases
and providing appropriate surveillance requirements to assure system
operability. A review of the performance history of the Standby Gas
Treatment System and licensing basis assumptions has determined that
the proposed changes do not adversely affect plant safety. Changes to
the SGTS performance requirements provide greater assurance of SGTS
operability. The proposed change for the completion time to place the
plant in a cold shutdown condition if limiting conditions for operation
can not be satisfied is consistent with the time frame specified in the
current specification and is consistent with Standard Technical
Specifications. The proposed amendment will not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW, Washington, DC 20037.
NRC Project Director: Ledyard B. Marsh.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: February 12, 1993, as supplemented by
letters dated August 20, 1993 and June 6, 1994.
Description of amendment request: This amendment request was
previously noticed in the Federal Register on April 14, 1993 (58 FR
19485). The June 6, 1994, submittal supplements the February 12, 1993,
application for amendment, and includes and incorporates the NRC staff
comments. The proposed amendment would modify the Technical
Specifications (TS) to implement the reactor coolant system (RCS) leak
before break (LBB) methodology detection criteria, in accordance with
the recommendations listed in Generic Letter (GL) 84-04.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes will require additional leak detection
instruments be operable to close Unresolved Safety Issue A-2,
``Asymmetrical Blowdown Loads on Reactor Primary Coolant System,'' for
the Fort Calhoun Station. Requiring additional instruments to be
operable does not increase the probability or consequences of an
accident since the safety function of the instruments is not being
altered.
The proposed changes require at least two different types of RCS
leak detection instruments, of diverse monitoring principles, be
operable or corrective actions be taken to restore the instrumentation
to operable status. Currently the Technical Specifications require only
one RCS leak detection instrument to be operable.
The probability of leaks occurring due to thermal or normal fatigue
is not affected as indicated in the fracture mechanics analysis
referenced in Generic Letter 84-04. No changes are proposed to primary
RCS piping systems or supports as a result of the proposed revision.
The proposed changes will ensure that a potential significant failure
does not go undetected within the Regulatory Guide 1.45 criteria as
noted in Generic Letter 84-04.
The Loss of Coolant Accident (LOCA) analysis will not be impacted
by the proposed change. The results of the current Fort Calhoun LOCA
analyses cited in Section 14.15 of the Updated Safety Analysis Report
(USAR) will not be impacted as a result of these changes.
(2) Create the possibility of a new or different kind of accident
from any previously analyzed.
It has been determined that a new or different kind of accident
will not be created due to the proposed changes since no new or
different modes of operation are created by this change. The existing
operating procedures were established to support an enhanced RCS leak
detection program. Operation of RCS leak detection instruments will not
differ from existing conditions.
(3) Involve a significant reduction in a margin of safety.
The margin of safety as defined in the basis for the Technical
Specifications is not changed or reduced by this proposed change.
Defining adequate RCS LBB monitoring is required to meet
recommendations provided in Generic Letter 84-04.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102.
Attorney for licensee: LeBoeuf, Lamb, Leiby, and MacRae, 1875
Connecticut Avenue NW., Washington, DC 20009-5728.
NRC Project Director: William D. Beckner.
Philadelphia Electric Company, Docket No. 50-352, Limerick Generating
Station, Unit 1, Montgomery County, Pennsylvania
Date of amendment request: June 6, 1994.
Description of amendment request: The amendment would remove the
controls for a remote shutdown system control valve and delete the
isolation signal for certain primary containment isolation valves from
TS Tables 3.3.7.4-1 and 3.6.3-1 respectively, as a result of
eliminating the steam condensing mode of the Redidual Heat Removal
system.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed Technical Specifications (TS) changes do not
involve a significant increase in the probability or consequences of an
accident previously evaluated.
The RHR system steam condensing mode is a non-safety related
function of the RHR system and has been eliminated at Limerick
Generating Station, Unit 1. These proposed changes will not affect any
components required to perform the safety-related function of the RHR
system.
The ability of the RHR system to respond to an accident will not be
degraded by the proposed changes. Valve HV-51-1F011A is locked closed
with the electrical power removed. The valve's handswitch which is part
of the remote shutdown panel (RSP) controls, does not perform any
function and will be physically removed from the RSP. The deletion of
the isolation signal for valves HV-C-51-1F103A and HV-C-51-1F104B will
not affect the ability of these valves to function as primary
containment isolation valves (PCIVs), since they are locked closed
already in their safety-related position, providing containment
isolation as manual PCIVs. Therefore, the proposed TS changes do not
involve an increase in the probability or consequences of an accident
previously evaluated.
2. The proposed TS changes do not create the possibility of a new
or different kind of accident from any accident previously evaluated.
No new failure modes of RHR system are created by the proposed TS
changes. All valves associated with the proposed changes are dedicated
specifically for the RHR system steam condensing mode, and will not
impact the operation of any components or piping required for other
modes of operation of the RHR system. These valves are locked-closed in
their safety-related position with the electrical power removed.
Therefore, the proposed TS changes do not create the possibility of a
new or different kind of accident from any previously evaluated.
3. The proposed TS changes do not involve a significant reduction
in a margin of safety.
The steam condensing mode is a non-safety related function of the
RHR system and, therefore, is not addressed in the TS. The controls for
remote shutdown system control valve HV-51-1F011A are not being used.
Presently, the valve is locked closed with the electrical power removed
and the valve's handswitch will be removed from the RSP, since it does
not perform any function. The proposed changes will not impact the safe
operation of LGS Unit 1. The deletion of the isolation signal for
valves HV-C-51-1F103A and HV-C-51-1F104B will not affect the ability of
these valves to function as primary containment isolation valves
(PCIVs), since they are locked closed already in their safety-related
position. Therefore, the proposed TS changes do not involve a reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101.
NRC Project Director: Charles L. Miller.
Philadelphia Electric Company, Docket No. 50-352, Limerick Generating
Station, Unit 1, Montgomery County, Pennsylvania
Date of amendment request: June 10, 1994.
Description of amendment request: The amendment involves a one-time
change affecting the Allowed Outage Time (AOT) for the Emergency
Service Water (ESW) System, Residual Heat Removal Service Water (RHRSW)
System, Suppression Pool Cooling, Suppression Pool Spray, and Low
Pressure Coolant Injection modes of the Residual Heat Removal System,
and Core Spray System to be extended from 3 and 7 days to 14 days
during the Limerick Generating Station (LGS), Unit 2 third refueling
outage scheduled to begin in January 1995. This proposed extended AOT
would allow adequate time to install isolation valves and cross-ties on
the ESW and RHRSW Systems to facilitate future inspections or
maintenance.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed Technical Specifications changes do not involve a
significant increase in the probability or consequences of an accident
previously evaluated.
The proposed one-time TS changes will not increase the probability
of an accident since it will only extend the time period that the `B'
ESW and RHRSW loops and the affected equipment can be out-of-service.
The extension of the time duration that certain equipment is out-of-
service has no direct physical impact on the plant. The proposed
inoperable systems are normally in a standby mode while the unit is in
OPCON 1 or 2 and are not directly supporting plant operation.
Therefore, they can have no impact on the plant that would make an
accident more likely to occur due to their inoperability.
During transients or events which require these systems to be
operating, there is sufficient capacity in the operable loops to
support plant operation or shutdown, in-so-much that failures that are
accident initiators will not occur more frequently than previously
postulated.
In addition, the consequences of an accident previously evaluated
in the SAR will not be increased. With the `B' loops of ESW and RHRSW
inoperable, a known quantity of equipment is either inoperable or the
equipment is not fully capable of fulfilling its design function under
all design conditions due to certain support systems not being
operable. Based on the support functions of the ESW and RHRSW systems,
a review of the plant was performed to determine the impacts that the
inoperable ESW and RHRSW `B' loops would have on other systems. The
impacts were identified for each system, as discussed in the preceding
Safety Assessment, and it was determined whether there were any adverse
[effects] on the systems. It was then determined how the adverse
[effects] would impact each system's design basis and overall plant
safety. The consequences of any postulated accidents occurring on Unit
1 during this AOT extension was found to be bounded by the previous
analyses as described in the SAR.
The existing AOTs limit the amount of time that the plant can
operate with certain equipment inoperable, where single failure
criteria is still met. The minimum equipment required to mitigate the
consequences of an accident and/or safely shutdown the plant will be
operable or the plant will be shutdown. Therefore, by extending certain
AOTs and extending the assumptions concerning the combinations of
events and single failures for the longer duration of each extended
AOT, we conclude, based on the evaluations above, that at least the
minimum equipment required to mitigate the consequences of an accident
and/or safely shutdown the plant will still be operable during the
extended AOT. Therefore, the consequences of an accident previously
evaluated in the SAR will not be increased.
Therefore, these proposed one-time TS changes will not result in a
significant increase in the probability or consequences of an accident
previously evaluated.
2. The proposed TS changes do not create the possibility of a new
or different kind of accident from any accident previously evaluated.
The proposed one-time TS changes will not create the possibility of
a different type of accident since it will only extend the time period
that the `B' ESW and RHRSW loops and the affected equipment can be out-
of-service. The extension of the time duration that certain equipment
is out-of-service has no direct physical impact on the plant and does
not create any new accident initiators. The systems involved are either
accident mitigation systems, safe shutdown systems or systems that
support plant operation. All of the possible impacts that the
inoperable equipment may have on its supported systems were previously
analyzed in the SAR and are the basis for the present TS ACTION
statements and AOTs. The impact of inoperable support systems for a
given time duration was previously evaluated and any accident
initiators created by the inoperable systems was evaluated. The
lengthening of the time duration does not create any additional
accident initiators for the plant.
Therefore, the proposed one-time TS changes will not create the
possibility of a new or different type of accident from any accident
previously evaluated.
3. The proposed TS changes do not involve a significant reduction
in a margin of safety.
The ESW and RHRSW systems and their supported systems are designed
with sufficient independence and redundancy such that the removal from
service of a component/subsystem will not prevent the systems from
performing their required safety functions. Since removal of an ESW and
a RHRSW loop from service with one unit in operation and the other unit
in a refueling outage is allowed by the current Technical
Specifications, then the concern is the reduced margin of safety
incurred by extending the affected AOTs.
The present ESW and RHRSW AOT limits were set to ensure that
sufficient safety-related equipment is available for response to all
accident conditions and that sufficient decay heat removal capability
is available for a LOCA/LOOP on one unit and simultaneous safe shutdown
of the other unit. A slight reduction in the margin of safety is
incurred during the proposed extended AOT due to the increased risk
that an event could occur in a fourteen day period versus a three or
seven day period. This increased risk is judged to be minimal due to
the low probability of an event occurring during the extended AOT and
based on the following discussion of minimum ECCS/decay heat removal
requirements.
The reduction in the margin of safety is not significant since the
remaining operable ECCS equipment is adequate to mitigate the
consequences of any accident. This conclusion is based on the
information contained in documents NEDO-24708A and NEDC-30936-A. These
documents described the minimum requirements to successfully terminate
a transient or LOCA initiating event (with scram), assuming multiple
failures with realistic conditions and were used to justify certain TS
AOTs per UFSAR sections 6.3.1.1.2.o and 6.3.3.1. The minimum
requirements for short term response to an accident would be either one
LPCI pump or one Core Spray loop in conjunction with ADS, which would
be adequate to re-flood the vessel and maintain core cooling sufficient
to preclude fuel damage. For long term response, the minimum
requirements would be one loop of RHR for decay heat removal, along
with another low pressure ECCS loop. These minimum requirements will be
met since implementation of the proposed TS changes will require the
operability of HPCI, ADS, two LPCI subsystems (or one LPCI subsystem
and one RHR subsystem during decay heat removal) and one Core Spray
subsystem be maintained during the 14 day period.
In addition, measures will be taken prior to or during the proposed
extended AOT for those fire regions that rely on one or more safe
shutdown methods which would all be unable to safely shutdown the plant
with inoperable loops of the ESW and RHRSW systems or the inoperable
systems that ESW or RHRSW support. These measures will offset the
increased risk of a fire event occurring in the vulnerable areas,
during the fourteen day versus three day AOT period. Therefore, the
proposed extended AOT does not adversely affect the approved level of
fire protection as described in UFSAR Appendix 9A (Fire Protection
Evaluation Report).
A special procedure will be written to administratively control the
requirement to maintain the operability of specified components and
implementation of any appropriate compensatory measures which are
deemed necessary during the proposed AOT. In addition, operations
personnel are fully qualified by normal periodic training to respond to
and mitigate a Design Basis Accident, including the actions needed to
ensure decay heat removal while LGS Unit 1 and Unit 2 are in the
operational configurations described within this submittal.
Accordingly, procedures are already in place that cover safe plant
shutdown and decay heat removal for situations applicable to those in
the proposed AOTs.
A Probabilistic Safety Assessment (PSA) Study was performed for an
ESW and RHRSW loop being out-of-service for 14 days on an operating
unit. This analysis includes EDG D12 being aligned to `A' ESW and HPCI
and RCIC not requiring room cooling. No other deviations from the
bounding assumptions used in the base PSA model were made. The Core
Damage Frequency (CDF) increased by 2.7x10-6, from 5.11x10-6/
reactor-year to 7.8x10-6 /reactor-year. In absolute terms, this is
not a significant increase in risk. In addition, the modifications to
be installed during this proposed extended AOT will allow for future
maintenance and inspections to be performed on the ESW and RHRSW loops
without removing an entire loop from service, which will reduce risk in
the future. For example, if the ESW loop unavailability, due to testing
or maintenance, is reduced by half, the CDF will decrease by more than
four percent. It will also minimize the potential need for future AOT
extensions on these systems.
Therefore, the implementation of the proposed one-time TS changes
will not involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Attorney for licensee: J.W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101.
NRC Project Director: Charles L. Miller.
Philadelphia Electric Company, Public Service Electric and Gas Company,
Delmarva Power and Light Company, and Atlantic City Electric Company,
Dockets Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station,
Units Nos. 2 and 3, York County, Pennsylvania
Date of application for amendments: May 10, 1994.
Description of amendment request: The licensee is proposing to
revise the Technical Specifications (TS) requirements governing the
minimum low pressure cooling availability when irradiated fuel is in
the reactor vessel and the reactor is in the cold condition.
Specifically, the proposed changes are: (1) Revise the TS section
titles in the Table of Contents to agree with the TS section titles in
the body of the TS. (2) Revise TS Section 3.5.A.1 to provide proper
reference to the revised TS Section 3.5.F. (3) Revise TS Section
3.5.A.3 to provide proper reference to the revised TS Section 3.5.F.
(4) Revise TS Section 3.5.B.1 to delete the reference to TS Section
3.5.F.3. (5) Revise TS Section 3.5.F to require the limiting condition
for operation (LCO) governing minimum low pressure cooling availability
when irradiated fuel is in the reactor vessel and the reactor is in the
cold condition to be identical with the corresponding LCO in NUREG-
1433, ``Standard Technical Specifications General Electric Plants, BWR/
4.'' (6) Revise TS Section 4.5.F to require the surveillance
requirements (SR) governing minimum low pressure cooling availability
when irradiated fuel is in the reactor vessel and the reactor is in the
cold condition to be identical with the corresponding SR in NUREG-1433.
(7) Revise TS BASES 3.5.A to delete the reference to the core spray
subsystem as also providing a source for flooding of the core in case
of accidental draining because the information is being added to TS
BASES 3.5.F. (8) Revise TS BASES 3.5.F to be consistent with the
corresponding TS BASES in NUREG-1433. (9) Revise TS BASES 4.5 to be
consistent with the corresponding TS BASES in NUREG-1433. (10) Revise
TS Section 3.7.A.1 to provide proper reference to the revised TS
Section 3.5.F.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Proposed changes 1, 2, 3, 4, 7, 8, 9 and 10 are administrative in
nature and involve no technical changes to the TS. These proposed
changes do not impact initiators of analyzed events or the assumed
mitigation of accidents or transients events. Therefore, these changes
do not involve an increase in the probability or consequences of an
accident previously evaluated.
Proposed changes 5 and 6 will not increase the probability of
initiating an analyzed event or alter assumptions relative to
mitigation of an accident or transient event. These changes will not
alter the operation of process variables or systems, structures, or
components (SSC) as described in the safety analyses. These changes do
not involve any physical changes to plant SSC. TS requirements that
govern Operability or routine testing and verification of plant
components and variables are not assumed to be initiators of any
analyzed event. The proposed changes will not alter the operation of
equipment assumed to be available for the mitigation of accidents or
transients by the plant safety analysis or licensing basis. The
proposed changes establish or maintain adequate assurance that
components are operable when necessary for the prevention or mitigation
of accidents or transients and that plant variables are maintained
within limits necessary to satisfy the assumptions for initial
conditions in the safety analysis. These changes have been confirmed to
ensure no previously evaluated accident has been adversely affected.
These changes will not allow continuous plant operation with plant
conditions during a unit outage such that a single failure will result
in a loss of any safety function. Therefore, the changes will not
involve a significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes do not involve a physical alteration of the
plant (no new or different type of equipment will be installed or
removed) and will not alter the method used by any system to perform
its design function. The proposed changes do not allow plant operation
in any mode that is not already evaluated. Therefore, these changes
will not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. The proposed changes do not involve a significant reduction in a
margin of safety.
Proposed changes 1, 2, 3, 4, 7, 8, 9 and 10 are administrative in
nature and will not involve any technical changes. These proposed
changes will not reduce a margin of safety because they have no impact
on any safety analysis assumptions. Because these changes are
administrative in nature, no question of safety is involved. Therefore,
these changes do not reduce the margin of safety.
Proposed changes 5 and 6 add some new requirements and make some
existing requirements more restrictive. These changes will not impact
any safety analysis assumptions. Adding new requirements and making
existing ones more restrictive either increases or does not affect the
margin of safety. As such, no question of safety is involved.
Therefore, these changes will not involve a significant reduction in a
margin of safety.
Proposed changes 5 and 6 also make three less restrictive changes.
The first change deletes the requirements for the containment cooling
system when the reactor is in the Cold Condition. The containment
cooling system is necessary to maintain primary containment Operable to
mitigate the release of radioactive material following a DBA [design
basis accident].
However, primary containment is not required to be Operable with
the reactor in the Cold Condition. As a result, the containment cooling
system is not needed to maintain the primary containment Operable with
the reactor in the Cold Condition. This change does not affect any
safety limits, operating limits, or design assumptions. This [change]
provides the benefit of allowing maintenance to be performed on the
containment cooling systems during a unit outage to ensure their
reliability during power operation. Therefore, this change does not
involve a significant reduction in a margin of safety.
The current TS requirement that both core spray systems and the
LPCI system be Operable during a refueling outage is being relaxed by
the second change to require one core spray subsystem and one LPCI
subsystem or two core spray subsystems to be Operable. This [change]
does not adversely affect any accident or transient analyses because
the change ensures adequate vessel inventory makeup is available in the
event of an inadvertent vessel draindown. The long term cooling
analysis following a design bases LOCA [loss of coolant accident]
demonstrates only one low pressure ECCS [emergency core cooling system]
injection/spray subsystem is required, post LOCA, to maintain the peak
cladding temperature below the allowable limit. This change will not
affect any safety limits, operating limits, or design assumptions. This
change provides the benefit of allowing maintenance to be performed on
the low pressure ECCS subsystems not required to be operable to ensure
their reliability during plant operation. Therefore, this change does
not involve a significant reduction in a margin of safety.
The final less restrictive change will allow low pressure
injection/spray subsystems to be inoperable during a refueling outage
if the spent fuel storage gates are removed and the water level is at
the required height over the top of the reactor pressure vessel flange.
This is acceptable because the water level requirement provides
sufficient coolant inventory to allow operator action to terminate any
inventory loss prior to fuel uncovery in the event of an inadvertent
draindown. This change will not affect any safety limits, operating
limits, or design assumptions. This change provides the benefit of
allowing maintenance to be performed on the low pressure ECCS
subsystems to ensure their continued reliability during plant
operation. Therefore, this change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
Attorney for Licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101.
NRC Project Director: Charles L. Miller.
Philadelphia Electric Company, Public Service Electric and Gas Company,
Delmarva Power and Light Company, and Atlantic City Electric Company,
Dockets Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station,
Units Nos. 2 and 3, York County, Pennsylvania
Date of application for amendments: June 9, 1994.
Description of amendment request: The licensee proposed the
following changes to its Technical Specifications (TS): (1) Revise TS
4.3.C.1 to require that each control rod be scram time tested after
each refueling outage or after a reactor shutdown that is greater than
120 days with reactor steam dome pressure greater than or equal to 800
psig prior to exceeding 40% of rated power. Scram time testing is not
required for control rods inserted per TS 3.3.B.1. (2) Replace TS
4.3.C.2 with the requirement to perform scram time testing with the
reactor steam dome pressure greater than or equal to 800 psig prior to
exceeding 40% of rated power for only those control rods associated
with the core cells affected by any fuel movement within the reactor
pressure vessel. (3) Add TS 4.3.C.3 to perform scram time testing for a
representative sample of control rods at least once per 120 days of
power operation with the reactor steam dome pressure greater than or
equal to 800 psig. (4) Add TS 4.3.C.4 to perform scram time testing at
any reactor steam dome pressure for individual control rods prior to
declaring them operable after work on the control rod or control rod
drive system is performed that could affect scram insertion time. (5)
Revise TS Bases 3.3.C and 4.3.C to describe: the rational for
performing scram time testing with reactor pressure greater than or
equal to 800 psig; the rationale for requiring control rods to be scram
time tested once per 120 days; what constitutes a representative sample
of control rods; examples of work that could affect scram times; and
the rational and methods for performing scram time testing following
work that could affect the scram insertion times. (6) Add TS 4.3.C.5 to
perform scram time testing with the reactor steam dome pressure greater
than or equal to 800 psig prior to exceeding 40% of rated power after
work on the control rod or control rod drive system that could affect
scram insertion time. (7) Revise TS 4.5.K.2 from performing scram time
testing of 19 or more control rods on a rotation basis to performing
scram time testing of a representative sample of control rods.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes will not involve any physical changes to plant
systems, structures, or components (SSC). These proposed changes will
not alter operation of process variables or SSC as described in the
safety analysis. The proposed changes establish or maintain adequate
assurance that components are operable when necessary for the
prevention or mitigation of accidents or transients and that plant
variables are maintained within limits necessary to satisfy the
assumptions for initial conditions in the safety analysis. In
particular, proposed change 1 is acceptable based on industry
experience with control rod scram time testing coupled with the
additional requirement in proposed change 4 that scram time testing of
any control rod on which work was performed must be satisfactorily
completed before that control rod can be declared operable. The
proposed changes will not allow continuous plant operation with plant
conditions such that a single failure will result in a loss of any
safety function. Therefore, the changes will not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
(2) The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes do not alter the plant configuration (no new
or different type of equipment will be installed or removed) and will
not alter the method used by any system to perform its design function.
The proposed changes do not allow plant operation in any mode that is
not already evaluated. Therefore, these changes will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
(3) The proposed changes do not involve a significant reduction in
a margin of safety.
Following a refueling outage, control rod scram time testing for
all control rods is currently required to be performed during
operational hydrostatic testing or during startup prior to
synchronizing the main turbine generator. Any control rods not tested
during operational hydrostatic testing must be tested at greater than
30% power but less than 40% power. Proposed change 1 will require that
scram time testing for all control rods be completed prior to exceeding
40% Reactor Power. This change is acceptable based on industry
experience with control rod scram time testing coupled with the
additional requirement in proposed change 4 that scram time testing of
any control rod on which work was performed must be satisfactorily
completed before that control rod can be declared operable. Proposed
changes 2, 3, 4 and 6 add some new requirements and make some existing
requirements more restrictive. The margin of safety is not reduced by
more restrictive changes. If anything, the margin of safety may
increase. Proposed change 5 revises the BASES to provide consistency
with the previously discussed SR [surveillance requirements] changes.
Proposed change 7 is administrative in nature and does not involve any
technical changes. Proposed changes 5 and 7 will not reduce a margin of
safety because they have no impact on any safety analysis assumptions.
Therefore, these changes will not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
Attorney for Licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101
NRC Project Director: Charles L. Miller.
Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: June 29, 1994.
Description of amendment request: The proposed changes would alter
the Plant Operating Review Committee's (PORC's) membership requirements
and would delegate a portion of PORC's procedure review
responsibilities for nuclear safety related procedures and procedure
changes to the line organizations. Section 6.5, ``Review and Audit,''
of the Technical Specifications (TSs) would be revised to modify the
composition of the PORC and delete review and audit responsibilities
for the Emergency and Security Plans from the TSs. The review and audit
responsibilities would be relocated to the respective Emergency and
Security Plans consistent with Generic Letter 93-07, ``Modification of
the Technical Specification Administrative Control Requirements for
Emergency and Security Plans.'' The proposed changes would also revise
Section 6.5 and Section 6.8, ``Procedures,'' of the TSs to delegate a
portion of the PORC's procedure review responsibilities for nuclear
safety related procedures to the line organizations. The PORC would
continue to perform safety reviews associated with procedures that are
of safety significance.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Consistent with the criteria of 10 CFR 50.92, the enclosed
application is judged to involve no significant hazards based on the
following information:
1. Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident previously
analyzed?
Response: The proposed changes do not involve a significant
increase in the probability or consequences of an accident previously
evaluated since: (1) PORC will continue to review environmental impact
and 10 CFR 50.59 safety evaluations associated with procedures and
procedure changes; (2) Only personnel knowledgeable in the affected
functional areas will review procedures and procedure changes; (3)
Review and approval personnel (designated technical reviewers,
qualified safety reviewers and responsible procedure owners) will be
identified in the appropriate administrative procedures; (4) Designated
technical reviewers shall meet or exceed the qualifications described
in Section 4 of ANSI N18.1-1971 [* * *] for applicable positions and
are designated by the Department Managers; 5) The designated technical
reviewers will be responsible for identifying whether additional cross
disciplinary reviews are required; (6) The qualified safety reviewers
will be responsible for reviewing the procedure changes from a safety
perspective; (7) The responsible procedure owners are designated by the
Resident Manager; and (8) The responsible procedure owners will be
responsible for verifying that procedure reviews are performed in
accordance with the administrative procedure governing the procedure
review and approval process.
The proposed changes (1) will add more detailed requirements
regarding procedure review and approval to the Technical Specifications
which will strengthen the controls over the process, and (2) will free
PORC from reviewing items that are outside the charter of a ``safety
review'' committee [* * *] [because non-safety significant items can
reduce the time that PORC members can spend on matters that are safety
significant. The proposed Technical Specification change establishes a
highly structured review and approval program for procedures.]
The proposed change to the PORC membership requirements would not
significantly increase the probability or consequences of an accident
because it does not adversely affect the level of expertise applied to
the PORC review function or its effectiveness. The PORC quorum is
currently composed of five members including up to two designated
alternates and a Chairman. [* * *] [This composition is not changed by
the proposed amendment.]
The miscellaneous administrative changes not related to the
procedure review and approval process or the PORC membership
requirements cannot affect the probability or consequences of an
accident because they do not affect plant operations[,] [* * *]
equipment[, or any safety-related activity.]
2. Does the proposed license amendment create the possibility of a
new or different kind of accident from any accident previously
evaluated?
Response: [* * *] [No physical changes to the plant or changes in
plant equipment operating procedures are being proposed.] The changes
are administrative and will not have any direct effect on equipment
important to safety. Changing the process by which procedures are
reviewed and approved cannot in itself create the possibility of a new
or different kind of accident. Furthermore, a documented safety review,
utilizing screening criteria, will be performed for all nuclear safety
related procedures and procedure changes. The proposed change
establishes detailed controls while allowing PORC to spend more time on
safety significant issues.
The proposed change to the PORC membership requirements would not
create the possibility of a new or different kind of accident from any
previously evaluated since no physical alterations of plant
configuration or changes to setpoints or operating parameters are
proposed.
The miscellaneous administrative changes not related to the
procedure review and approval or the PORC Membership requirements
cannot create the possibility of an accident because they do not affect
plant operations[,] [* * *] equipment [or any safety-related activity.]
3. Does the proposed amendment involve a significant reduction in a
margin of safety?
Response: The proposed amendment does not involve a significant
reduction in the margin of safety because a program controlled by
Administrative Procedures using designated technical reviewers approved
by the Department Managers will be in place to review new procedures
and procedure changes. A 10 CFR 50.59 screening of each new procedure
and permanent procedure change will be performed by a qualified safety
reviewer, and PORC will continue to review 10 CFR 50.59 Safety and
Environmental Impact Evaluations associated with procedures and
procedure changes. Cross disciplinary reviews will be conducted as
appropriate. Thus, the margin of safety will be maintained by
implementing the new procedure review and approval process.
The proposed change to the PORC membership requirements would not
involve a significant reduction in the margin of safety since the level
and quality of PORC review will be maintained and there will not be an
adverse change to the collective educational background and work
experience of PORC. [There will not be an adverse loss of PORC
effectiveness as a result of this change.] The PORC quorum is currently
composed of five members including up to two designated alternates and
a Chairman. [* * *] [This composition is not changed by the proposed
changes.]
The miscellaneous administrative changes not related to the
procedure review and approval process or PORC membership program cannot
reduce any margin of safety because they do not affect any safety
related activity or equipment. These changes increase the probability
that the Technical Specifications are correctly interpreted by
clarifying information.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10601.
Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle,
New York, New York 10019.
NRC Project Director: Michael L. Boyle.
Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: June 17, 1994.
Description of amendment request: The proposed changes would revise
Section 6.5, ``Review and Audit,'' of the Technical Specifications
(TSs) to modify the composition of the Plant Operating Review Committee
(PORC) and delete review and audit responsibilities for the Emergency
and Security Plans from the TSs. The review and audit responsibilities
would be relocated to the respective Emergency and Security Plans
consistent with Generic Letter 93-07, ``Modification of the Technical
Specification Administrative Control Requirements for Emergency and
Security Plans.'' The proposed changes would also revise Section 6.5
and Section 6.8, ``Procedures,'' of the TSs to delegate a portion of
the PORC's procedure review responsibilities for nuclear safety related
procedures to the line organizations. The PORC would continue to
perform safety reviews associated with procedures that are of safety
significance.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Operation of the FitzPatrick plant in accordance with the proposed
amendment would not involve a significant hazards consideration as
defined in 10 CFR 50.92, since the proposed changes would not:
1. involve a significant increase in the probability of an accident
or consequence previously evaluated.
The proposed changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated since:
1) PORC will continue to review environmental impact and 10 CFR 50.59
safety evaluations associated with procedures and procedure changes; 2)
Only personnel knowledgeable in the affected functional areas will
review procedures and procedure changes; 3) Review and approval
personnel (designated technical reviewers, and responsible procedure
owners) will be identified in the appropriate administrative
procedures. Designated technical reviewers shall meet or exceed the
qualifications described in section 4 of ANSI N18.1-1971 for applicable
positions. Designated technical reviewers are designated by the
Department Managers. The responsible procedure owners are designated by
the Resident Manager; 4) The designated technical reviewers will be
responsible for identifying whether additional cross-disciplinary
reviews are required; 5) The qualified safety reviewers will be
responsible for reviewing the procedure changes from a safety
perspective; and 6) The responsible procedure owners will be
responsible for verifying that procedure reviews are performed in
accordance with the administrative procedure governing the procedure
review and approval process.
The proposed changes (1) will add more detailed requirements
regarding procedure review and approval to the Technical Specifications
which will strengthen the controls over the process, and (2) will free
PORC from reviewing items that are outside the charter of a ``safety
review'' committee because non-safety significant items can reduce the
time that PORC members can spend on matters that are safety
significant. The proposed Technical Specification change establishes a
highly structured review and approval program for procedures.
The proposed change to the PORC Membership requirements would not
significantly increase the probability or consequences of an accident
because it does not adversely affect the level of expertise applied to
the PORC review function. There will not be an adverse loss of PORC
effectiveness as a result of this change. The PORC quorum is currently
composed of five members including up to two designated alternates and
a Chairman. This composition is not changed by the proposed amendment.
The miscellaneous administrative changes not related to the
procedure review and approval process or the PORC Membership
requirements cannot affect the probability or consequences of an
accident because they do not affect operations, equipment, or any
safety-related activity.
2. Create the possibility of a new or different kind of accident
from those previously evaluated.
No physical changes to the plant or changes in plant equipment
operating procedures are being proposed. The changes are administrative
and will not have any direct effect on equipment important to safety.
Changing the process by which procedures are reviewed and approved
cannot in itself create the possibility of a new or different kind of
accident. Furthermore, a documented safety review, utilizing screening
criteria, will be performed for all nuclear safety related procedures
and procedure changes. The proposed change establishes detailed
controls while allowing PORC to spend more time on safety significant
issues.
The proposed change to the PORC Membership requirements would not
create the possibility of a new or different kind of accident from any
previously evaluated since no physical alterations of plant
configuration or changes to setpoints or operating parameters are
proposed.
The miscellaneous administrative changes not related to the
procedure review and approval or the PORC Membership requirements
cannot create the possibility of an accident because they do not affect
operations, equipment or any safety-related activity.
3. Involve a significant reduction in the margin of safety.
The proposed amendment does not involve a significant reduction in
the margin of safety because a program controlled by Administrative
Procedures using designated technical reviewers approved by the
Department Managers will be in place to review new procedures and
procedure changes. A 10 CFR 50.59 screening of each new procedure and
permanent procedure change will be performed by a qualified safety
reviewer and PORC will continue to review 10 CFR 50.59 Safety and
Environmental Impact Evaluations associated with procedures and
procedure changes. Cross-disciplinary reviews will be conducted as
appropriate. Thus the margin of safety will be maintained by
implementing the new procedure review and approval process.
The proposed change to the PORC Membership requirements would not
involve a significant reduction in the margin of safety since the level
and quality of PORC review will be maintained and there will not be an
adverse change to the collective educational background and work
experience of PORC. There will not be an adverse loss of PORC
effectiveness as a result of this change. The PORC quorum is currently
composed of five members including up to two designated alternates and
a Chairman. This composition is not changed by the proposed changes.
The miscellaneous administrative changes not related to the
procedure review and approval process or PORC Membership program cannot
reduce any margin of safety because they do not affect any safety-
related activity or equipment. These changes increase the probability
that the Technical Specifications are correctly interpreted by
clarifying information.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, New York 10019.
NRC Project Director: Michael L. Boyle.
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New
Jersey
Date of amendment requests: February 18, 1994, as supplemented by
letter dated April 6, 1994 for Salem Unit 1; and March 28, 1994 for
Salem Unit 2.
Description of amendment request: The proposed change to Salem Unit
1 Technical Specifications (TS) replaces the main feedwater control and
control bypass valves with the main feedwater stop check valves for the
Containment Isolation Function. The proposed change to Salem Unit 2 TS
adds a footnote to the 21-24 BF22 (main feedwater stop check valves) on
Table 3.6-1, ``Containment Isolation valves.'' This note identifies
those containment isolation valves that are not subject to 10 CFR 50
Appendix J, Type C leakage testing.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below.
1. Do not involve a significant increase in the probability or
consequence of an accident previously evaluated.
The Salem Unit 1 main feedwater stop check valves provide the same
isolation function presently accomplished by the main feedwater control
and control bypass valves, without reliance on an actuation signal.
Positive closure is assured during all postulated accident scenarios,
through remote-manual controls in the main control room. These valves
satisfy the requirements of GDC 57 for Containment Isolation.
A previous amendment request for Salem Unit 2 neglected to
designate the main feedwater stop check valves as exempt from Type C
leakage testing. That request was subsequently approved as Salem Unit 2
Amendment 128. The amended Technical Specifications are now
inconsistent with the Salem Updated Final Safety Analysis Report, which
correctly shows that these valves are exempt from Type C leakage
testing. Valve functionality and operation are not affected by this
change.
Therefore, it may be concluded that the proposed changes do not
involve a significant increase in the probability or consequences of an
accident previously evaluated.
2. Do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
The Salem Unit 1 main feedwater stop check valves were originally
intended to perform the Containment Isolation Function. The only
changes to the original plant design were the addition of motor
operators and the upgrading of associated controls to safety-related.
These changes bring the stop check valves into compliances with GDC 57
requirements, and ensure positive valve closure during all postulated
accident scenarios. As stated above, a previous amendment request for
Salem Unit 2 neglected to designate the main feedwater stop check
valves as exempt from Type C leakage testing. The Salem Updated Final
Safety Analysis Report correctly shows that these valves are exempt
from Type C leakage testing. Valve functionality and operation are not
affected by these changes.
The changes do not involve modifications to plant equipment or
operation. Therefore, no new or different accident can be created by
these changes.
3. Do not involve a significant reduction in a margin of safety.
Check valves provide inherent isolation from reverse flow
conditions. Stop check valves provide increased safety due to the
positive closure feature. Motor operators with remote-manual closure
capability, allow positive closure from the main control room during
all postulated accident scenarios. These features ensure that an
adequate margin of safety is maintained.
Additionally, feedwater isolation, utilizing the main feedwater
control and control bypass valves, occurs through Reactor Trip and/or
Engineered Safety Features actuation. This feature is unaffected by the
proposed changes and redundant to the stop check valves for Containment
Isolation. There are no modifications to plant equipment or operation
involved. Feedwater system operation during normal and accident
conditions remains the same.
Therefore, it may be concluded that the proposed changes do not
involve a significant reduction in a margin of safety.
Based on this review, it appears that the three standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment requests involve no significant hazards
consideration.
Local Public Document Room location: Salem Free Public library, 112
West Broadway, Salem, New Jersey 08079.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: Charles L. Miller.
Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of amendment request: May 13, 1994, as supplemented June 24,
1994.
Description of amendment request: The proposed amendment would
revise the Ginna Station Technical Specifications (TSs) Section 6.0
``Administrative Controls,'' to be consistent with the criteria
contained in the NRC Final Policy Statement of Technical Specifications
Improvements for Nuclear Power Reactors, and NUREG-1431 ``Standard
Technical Specifications, Westinghouse Plants,'' September 1993. The
proposed amendment would also relocate to other programs and documents,
several TS requirements in accordance with this criteria.
The May 13, 1994, request supersedes the request of March 23, 1992,
published in the Federal Register on November 25, 1992 (57 FR 55589).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Operation of Ginna Station in accordance with the proposed
changes does not involve a significant increase in the probability or
consequences of an accident previously evaluated. The changes are
consistent with the Final Policy Statement on Technical Specifications
Improvements for Nuclear Power Reactors and NUREG-1431 and have
therefore, been previously evaluated by the NRC. Implementation of
these changes is expected to result in a significant human factors
improvement and enable RG&E [Rochester Gas & Electric] and the NRC to
focus on the most important requirements without any reduction in
safety. The changes which do not duplicate NRC guidance in NUREG-1431
are currently addressed by existing technical specifications and
regulations, the Ginna Station license, plant procedures, or the QA
[Quality Assurance] program.
2. Operation of Ginna Station in accordance with the proposed
changes does not create the possibility of a new or different kind of
accident from any accident previously evaluated. The Administrative
Controls section contains those requirements that are not covered by
other technical specifications which are considered necessary to assure
safe operation of the facility. The majority of changes are consistent
with the Final Policy Statement on Technical Specifications
Improvements for Nuclear Power Reactors and NUREG-1431 and have
therefore, been previously evaluated by the NRC. [* * *].
3. Operation of Ginna Station in accordance with the proposed
changes does not involve a significant reduction in a margin of safety.
All requirements removed from technical specifications are relocated to
other programs and documents. These alternative programs and documents
are controlled by existing regulations which provide a more appropriate
vehicle for addressing changes and compliance. There were no
administrative control requirements which were removed from technical
specifications and not addressed by other regulations. Therefore, there
is no significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Rochester Public Library, 115
South Avenue, Rochester, New York 14610.
Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400
L Street, NW., Washington, DC 20005.
NRC Project Director: Walter R. Butler.
Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant,
Unit 2, Hamilton County, Tennessee
Date of amendment request: June 28, 1994 (TS 94-09).
Description of amendment request: The proposed amendment would
revise the Sequoyah Nuclear Plant Unit 2 Technical Specifications (TS)
surveillance requirements, bases, and a Limiting Condition For
Operation to incorporate alternate steam generator tube plugging
criteria at tube support plate intersections. The proposed changes
would be implemented for Fuel Cycle 7 only and would affect: (1) TS
4.4.5.2.c.2 to address bobbin probe inspections; (2) TS 4.4.5.3.d to
address future inspections of tubes where the interim criteria is used;
(3) TS 4.4.5.4.a.6 to address application of the interim criteria for
indications found within the thickness of the tube support plate; (4)
TS 4.4.5.4.a.10 to address application of the tube plugging alternate
criteria and evaluation of indications; (5) TS 4.4.5.5.d and 4.4.5.5.e
to address reporting requirements and information to be reported to the
Commission regarding application of the criteria; (6) TS 3.4.6.2.c to
reduce the allowable reactor coolant system total primary-to-secondary
leakage through all steam generators from 1 gallon per minute to 600
gallons per day and from any one steam generator from 500 gallons per
day to 150 gallons per day; and (7) Bases 3/4.4.5 and 3/4.4.6.2 to
reflect the new primary-to-secondary leakage limits and add a
reference.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
TVA has evaluated the proposed TS change and has determined that it
does not represent a significant hazards consideration based on
criteria established in 10 CFR 50.92(c). Operation of Sequoyah Nuclear
Plant (SQN) in accordance with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Testing of model boiler specimens for free-span tubing (no tube
support plate restraint) at room temperature conditions shows burst
pressures in excess of 500 pounds per square inch (psi) for indications
of outer-diameter stress corrosion cracking with voltage measurements
as high as 19 volts. Burst testing performed on intersections pulled
from SQN with up to a 1.9-volt indication shows measured burst pressure
in excess of 6,600 psi at room temperature. Burst testing performed on
pulled tubes from other plants with up to 7.5-volt indications shows
burst pressures in excess of 6,300 psi at room temperatures. Correcting
for the effects of temperature on material properties and minimum
strength levels (as the burst testing was done at room temperature),
tube burst capability significantly exceeds the safety-factor
requirements of NRC Regulatory Guide (RG) 1.121.
Tube burst criteria are inherently satisfied during normal
operating conditions because of the proximity of the tube support plate
(TSP). Test data indicates that tube burst cannot occur within the TSP,
even for tubes that have 100 percent throughwall electrodischarge
machining notches, 0.75-inch long, provided that the TSP is adjacent to
the notched area. Since tube-to-tube support plate proximity precludes
tube burst during normal operating conditions, use of the criteria must
retain tube integrity characteristics that maintain a margin of safety
of 1.43 times the bounding faulted condition steam line break (SLB)
pressure differential. During a postulated SLB, the TSP has the
potential to deflect during blowdown following a main SLB, thereby
uncovering the TSP intersections.
Based on the existing database, the RG 1.121 criterion requiring
maintenance of a safety factor of 1.43 times the SLB pressure
differential on tube burst is satisfied by 7/8-inch-diameter tubing
with bobbin coil indications with signal amplitudes less than 8.82
volts, regardless of the indicated depth measurement. A 2.0-volt
plugging criterion (resulting in a projected end-of-cycle {EOC}
voltage) compares favorably with the 8.82-volt structural limit
considering the extremely slow apparent voltage growth rates and few
numbers of indications at SQN. Using the established methodology of RG
1.121, the structural limit is reduced by allowances for uncertainty
and growth to develop a beginning of cycle (BOC) repair limit that
would preclude indications at EOC conditions that exceed the structural
limit. The nondestructive examination (NDE) uncertainty component is
20.5 percent, and is based on the Electric Power Research Institute
(EPRI) alternate repair criteria (ARC).
Because of the few number of indications at SQN, the EPRI
methodology of applying a growth component of 35 percent per effective
full power year (EFPY) will be used. Near-term operating cycles at SQN
are expected to be bounded by 1.23 years, therefore, a 43 percent
growth component is appropriate. When these allowances are added to the
BOC interim plugging criteria of 2.0 volts in a deterministic bounding
EOC voltage of approximately 3.26 volts for Cycle 7, operation can be
established. A 5.56-volt deterministic safety margin exists (8.82
structural limit--3.26-volt EOC equal 5.56-volt margin).
For the voltage/burst correlation, the EOC structural limit is
supported by a voltage of 8.82 volts. Using this structural limit of
8.82 volts, a BOC maximum allowable repair limit can be established
using the guidance of RG 1.121. The BOC maximum allowable repair limit
should not permit the existence of EOC indications that exceed the
8.82-volt structural limit. By adding NDE uncertainty allowances and an
allowance for crack growth to the repair limit, the structural limit
can be validated. Therefore, the maximum allowable BOC repair limit
(RL) based on the structural limit of 8.82 volts can be represented by
the expression:
RL + (0.205 x RL) + (0.43 x RL) = 8.82 volts, or,
the maximum allowable BOC repair limit can be expressed as,
RL = 8.82-volt structural limit/1.64 = 5.4 volts.
It is reasonable that this RL (5.4 volts) could be applied for
interim plugging criteria (IPC) implementation to repair bobbin
indications greater than 2.0 volts independent of rotating pancake coil
(RPC) confirmation of the indication. Conservatively, an upper limit of
3.6 volts will be used to assess tube integrity for those bobbin
indications that are above 2.0 volts but do not have confirming RPC
calls. This 3.6-volt upper limit for nonconfirmed RPC calls is
consistent with other recently approved IPC programs.
The conservatism of the growth allowance used to develop the repair
limit is shown by the most recent SQN eddy current data. Two tubes
plugged in Unit 1 during the last outage had less than one volt of
growth over the past five operating cycles. Only one tube in Unit 2
required repair because of outer-diameter stress corrosion (ODSCC) at
the TSP intersections.
Relative to the expected leakage during accident condition
loadings, it has been previously established that a postulated main SLB
outside of containment, but upstream of the main steam isolation valve
(MSIV), represents the most limiting radiological condition relative to
the IPC. In support of implementation of the IPC, it will determine
whether the distribution of cracking indications at the TSP
intersections at the end of Cycle 7 for Unit 2 is projected to be such
that primary-to-secondary leakage would result in site boundary doses
within a small fraction of the 10 CFR 100 guidelines. A separate
analysis has determined this allowable SLB leakage limit to be 4.3
gallons per minute (gpm) in the faulted loop. This limit uses the TS
reactor coolant system (RCS) Iodine-131 activity level of 1.0
microcuries per gram dose equivalent Iodine-131 and the recommended
Iodine-131 transient spiking values consistent with NUREG-0800. The
projected SLB leakage rate calculation methodology prescribed in
Section 3.3 of draft NUREG-1477 is used to calculate EOC leakage.
Because of the relatively low number of indications at SQN, it is
expected that the actual leakage values will be far less than this
limit. Additionally, the current Iodine-131 levels at SQN range from
about 25 to 100 times less than the TS limit of 1.0.
Application of the criteria requires the projection of postulated
SLB leakage, based on the projected EOC voltage distribution for Cycle
7. Projected EOC voltage distribution is developed using the most
recent EOC eddy current results and a voltage measurement uncertainty.
Data indicates that a threshold voltage of 2.8 volts would result in
throughwall cracks long enough to leak at SLB condition. Draft NUREG-
1477 requires that all indications to which the IPC are applied must be
included in the voltage projection. Tube pull results from another
plant with 7/8-inch tubing with a substantial voltage growth database
have shown that tube wall degradation of greater than 40 percent
throughwall was readily detectable either by the bobbin or RPC probe.
The tube with the maximum throughwall penetration of 56 percent (42
average) had a voltage of 2.02 volts. This indication also was the
largest recorded bobbin voltage from the EOC eddy current data. Based
on the SQN pulled tube and industry pulled tube data supporting a lower
threshold for SLB leakage of 2.8 volts, inclusion of all IPC
intersections in the leakage model is quite conservative. The ODSCC
occurring at SQN is in its earliest stages of development. The
conservative bounding growth estimations to be applied to the expected
small number of indications for the upcoming inspection should result
in very small levels of predicted SLB leakage. Historically, SQN has
not identified ODSCC as a contributor to operational leakage. The
current leakage level at SQN is less than 1.0 gallon per day (gpd).
In order to assess the sensitivity of an indication's BOC voltage
to EOC leakage potential, a Monte Carlo simulation was performed for a
2.0-volt BOC indication. The maximum EOC voltage (at 99.8 percent
cumulative probability) was found to be 4.8 volts. Using NUREG-1477 and
EPRI leakage models, the leakage component from an indication of this
magnitude is 0.12 and 0.028 gpm, respectively.
Therefore, as implementation of the 2.0-volt IPC criterion during
Cycle 7 in Unit 2 does not adversely affect steam generator (S/G) tube
integrity and implementation will be shown to result in acceptable dose
consequences, the proposed amendment does not result in any increase in
the probability or consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
Implementation of the proposed S/G tube IPC criteria does not
introduce any significant changes to the plant design basis. Use of the
criteria does not provide a mechanism that could result in an accident
outside of the region of the TSP elevations; no ODSCC is occurring
outside the thickness of the TSP. Neither a single or multiple tube
rupture event would be expected in a S/G in which the plugging criteria
have been applied (during all plant conditions).
TVA will implement a maximum leakage rate limit of 150 gpd per S/G
to help preclude the potential for excessive leakage during all plant
conditions. The SQN TS limits on primary-to-secondary leakage at
operating conditions are to be a maximum of 0.42 gpm (600 gpd) for all
S/Gs, or, a maximum of 150 gpd for any one S/G. The RG 1.121 criterion
for establishing operational leakage rate limits that require plant
shutdown is based upon leak-before-break considerations to detect a
free-span crack before potential tube rupture during faulted plant
conditions. The 150-gpd limit should provide for leakage detection and
plant shutdown in the event of the occurrence of an unexpected single
crack resulting in leakage that is associated with the longest
permissible crack length. RG 1.121 acceptance criteria for establishing
operating leakage limits are based on leak-before-break considerations
such that plant shutdown is initiated if the leakage associated with
the longest permissible crack is exceeded. The longest permissible
crack is the length that provides a factor of safety of 1.43 against
bursting at faulted conditions maximum pressure differential. A voltage
amplitude of 8.82 volts for typical ODSCC corresponds to meeting this
tube burst requirement at a lower 95 percent prediction limit on the
burst correlation coupled with 95/95 lower tolerance limit material
properties. Alternate crack morphologies can correspond to 8.82 volts
so that a unique crack length is not defined by the burst pressure
versus voltage correlation. Consequently, typical burst pressure versus
through-wall crack length correlations are used below to define the
``longest permissible crack'' for evaluating operating leakage limits.
The single through-wall crack lengths that result in tube burst at
1.42 times the SLB pressure differential and the SLB pressure
differential alone are approximately 0.57 inch and 0.84 inch,
respectively. A leak rate of 150 gpd will provide for detection of 0.4-
inch-long cracks at nominal leak rates and 0.6-inch-long cracks at the
lower 95 percent confidence level leak rates. Since tube burst is
precluded during normal operation because of the proximity of the TSP
to the tube and the potential exists for the crevice to become
uncovered during SLB conditions, the leakage from the maximum
permissible crack must preclude tube burst at SLB conditions. Thus, the
150-gpd limit provides for plant shutdown before reaching critical
crack lengths for SLB conditions. Additionally, this leak-before-break
evaluation assumes that the entire crevice area is uncovered during
blowdown. Partial uncover will provide benefit to the burst capacity of
the intersection.
As S/G tube integrity upon implementation of the 2.0-volt IPC
continues to be maintained through in-service inspection and primary-
to-secondary leakage monitoring, the possibility of a new or different
kind of accident from any accident previously evaluated is not created.
3. Involve a significant reduction in a margin of safety. The use
of the voltage based bobbin probe interim TSP elevation plugging
criteria at SQN is demonstrated to maintain S/G tube integrity
commensurate with the criteria of RG 1.121. RG 1.121 describes a method
acceptable to the NRC staff for meeting General Design Criteria (GDC)
14, 15, 31, and 32 by reducing the probability or the consequences of
S/G tube rupture. This is accomplished by determining the limiting
conditions of degradation of S/G tubing, as established by in-service
inspection, for which tubes with unacceptable cracking should be
removed from service. Upon implementation of the criteria, even under
the worst-case conditions, the occurrence of ODSCC at the TSP
elevations is not expected to lead to a S/G tube rupture event during
normal or faulted plant conditions. The EOC distribution of crack
indications at the TSP elevations will be confirmed to result in
acceptable primary-to-secondary leakage during all plant conditions and
that radiological consequences are not adversely impacted.
In addressing the combined effects of loss-of-coolant accident
(LOCA), plus safe shutdown earthquake (SSE) on the S/G component (as
required by GDC 2), it has been determined that tube collapse may occur
in the S/Gs at some plants. This is the case as the TSP may become
deformed as a result of lateral loads at the wedge supports at the
periphery of the plate because of the combined effects of the LOCA
rarefraction wave and SSE loadings. Then, the resulting pressure
differential on the deformed tubes may cause some of the tubes to
collapse.
There are two issues associated with S/G tube collapse. First, the
collapse of S/G tubing reduces the RCS flow area through the tubes. The
reduction in flow area increases the resistance to flow of steam from
the core during a LOCA, which in turn, may potentially increase the
peak clad temperature (PCT). Second, there is a potential that partial
through-wall cracks in tubes could progress to through-wall cracks
during tube deformation or collapse.
Consequently, since the leak-before-break methodology is applicable
to the SQN reactor coolant loop piping, the probability of breaks in
the primary loop piping is sufficiently low that they need not be
considered in the structural design of the plant. The limiting LOCA
event becomes either the accumulator line break or the pressurizer
surge line break. LOCA loads for the primary pipe breaks were used to
bound the conditions at SQN for smaller breaks. The results of the
analysis using the larger break inputs show that the LOCA loads were
found to be of insufficient magnitude to result in S/G tube collapse or
significant deformation. The LOCA, plus SSE tube collapse evaluation
performed for another plant with Series 51 S/Gs using bounding input
conditions (large-break loadings), is considered applicable to SQN.
Addressing RG 1.83 considerations, implementation of the bobbin
probe voltage based interim tube plugging criteria of 2.0 volt is
supplemented by: (1) enhanced eddy current inspection guidelines to
provide consistency in voltage normalization, (2) a 100 percent eddy
current inspection sample size at the TSP elevations, and (3) RPC
inspection requirements for the larger indications left in service to
characterize the principal degradation as ODSCC.
As noted previously, implementation of the TSP elevation plugging
criteria will decrease the number of tubes that must be repaired. The
installation of S/G tube plugs reduces the RCS flow margin. Thus,
implementation of the alternate plugging criteria will maintain the
margin of flow that would otherwise be reduced in the event of
increased tube plugging.
Based on the above, it is concluded that the proposed license
amendment request does not result in a significant reduction in margin
of safety.
The NRC has reviewed the licensee's analysis and, based on this
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902.
NRC Project Director: Frederick J. Hebdon.
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: February 14, 1994, as supplemented by
letter dated May 17, 1994.
Brief description of amendments: The proposed changes revise the
Technical Specifications to allow power ascension above 50% rated
thermal power (RTP) with a quadrant power tilt ratio greater than 1.02
provided the assumptions of affected safety analyses are confirmed to
be satisfied.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not increase the probability or
consequences of a previously evaluated accident.
The proposed changes affect the Action Statements which are to be
taken when it is discovered that the Quadrant Power Tilt Ratio (QPTR)
is greater than the value specified in the Limiting Condition for
Operation. The frequency for determining QPTR has been reduced. The
requirements to reduce power to below 50% RTP and to reduce the power
range flux trip setpoints based on QPTR have been replaced by similar
requirements based on FQ(Z) and FNH. The
requirement to correct the cause prior to increasing power and
verifying QPTR hourly is replaced with specific requirements to verify
that FQ(Z) and FNH are within their limits, the
safety analyses remain valid, and the excore detectors are re-
normalized to indicate zero quadrant power tilt. If the proposed
actions are not met, a requirement to reduce power to 50%
RTP within 4 hours was added.
The only item above that could affect the probability of an
accident is the removal of the requirement to reduce the power range
neutron flux setpoints. However, because the Protection Cabinets must
be entered to make these adjustments, eliminating the requirement to
adjust these setpoints actually slightly reduces the probability of an
inadvertent plant trip. Thus, the changes do not increase the
probability of an accident previously evaluated and may reduce the
probability of a plant trip.
The proposed Action Statements, require that accident analyses be
re-evaluated to confirm that the results remain valid within 24 hours.
Prior to completion of this confirmation, the plant is not permitted to
operate at a power level higher than is permitted under the current
specification. If the re-evaluation of accident analyses cannot confirm
that the plant is within the accident analyses results, the required
actions are similar to the requirements of the current specification.
Although higher initial power levels generally increase accident
consequences, once the accident analyses are confirmed to be valid, the
consequences of any accident will be within analyzed acceptable limits.
Thus, the higher plant power levels permitted by the proposed changes
do not significantly increase the consequences of any accidents
previously evaluated.
The proposed specification does not require a reduction in Power
Range Neutron Flux--High reactor trip setpoints during the time the
appropriate peaking factor surveillances are being performed. The
interval during which the proposed specification permits operation
without reduced setpoints (and unverified peaking factors) is longer
than is permitted under the current specification. However, the
consequences of any accident which could occur during this interval are
the same as for the conditions prior to resetting the trip setpoints in
the current specification. Therefore the change does not increase the
consequences of any accident which could occur during this interval.
The impact of the extended interval is addressed in response to
question (3) below.
Based on the discussions above, the proposed changes do not involve
an increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new or
different kind of accident from any previously evaluated accident.
The proposed changes do not involve any hardware changes. System
operation has not been changed to create any new system configurations
which were not previously allowed. Therefore, the proposed changes do
not create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. The proposed change does not involve a significant reduction in
a margin of safety.
The limits for the parameters of concern (QPTR, FQ(Z) and
FNH) remain unchanged. The acceptance criteria for
analyzed events also remain the same. The margin of safety established
by the LCOs [Limiting Conditions for Operation] remains unchanged.
The only impact of the proposed changes is an increase in the
allowed duration of operation above 50% RTP without a reduction in the
Power Range Neutron Flux--High trip setpoint. This could potentially
affect a margin of safety by allowing operation at conditions which are
potentially outside the assumptions of the accident analyses for an
interval longer than is permitted under the current specification. The
impact on safety margin is not considered to be significant, however,
because: the allowed interval is still small (24 hours versus the
current 6 hours); the likelihood of an accident during the interval is
small; and, it is considered unlikely that the peaking factors would be
outside their limits without outer indications.
Thus, it is concluded that the proposed changes do not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 701 South Cooper, P.O.
Box 19497, Arlington, Texas 76019.
Attorney for licensee: George L. Edgar, Esq., Newman and
Holtzinger, 1615 L Street, NW., Suite 1000, Washington, DC 20036.
NRC Project Director: William D. Beckner.
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: March 28, 1994.
Brief description of amendments: The proposed amendment would
revise Section 6 of the Technical Specifications (TS) by deleting a
reference to a no longer used loss of coolant accident (LOCA) topical
report and adding a reference to a new steamline break methodology
topical report.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve an increase in the
probability or consequences of a previously evaluated accident.
The NRC assures that appropriate core operating limits are
established by requiring that they be determined using NRC approved
analytical methods. These approved methods are described in the
documents listed in TS Section 6.9.1.6b. TU Electric has developed the
analysis capability to evaluate the core operating limits. The
methodologies used by TU Electric have been documented in a series of
TU Electric submittals which were reviewed and approved by the NRC.
This TS revision adds the topical report which describes the TU
Electric steamline break analysis methodology to TS Section 6.9.1.6b.
Also, the Westinghouse report which describes the methodology
previously used in the analysis of Unit 1 large break LOCAs is no
longer used and is being deleted. Large break LOCA analyses for Unit 1
are now performed using NRC approved TU Electric methodology.
Because the revisions are administrative only, they cannot directly
affect the probability or the consequences of any previously evaluated
accident. The steamline break analysis methodology is part of a group
[of] methodologies which are authorized by the technical specifications
to be used to verify that each reload cycle continues to satisfy the
core operating limits. The core operating limits are set to assure that
relevant plant parameters are maintained such that potential accidents
are within the bounds of the accident analyses. Because the applicable
limits of the safety analyses will be verified to be satisfied using
authorized methodologies, there is no significant impact on the
consequences of an accident previously evaluated. In addition, since
the core operating limits do not affect any accident initiators, the
change has no impact on the probability of any accident previously
analyzed.
2.The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes involve a change in the permissible analysis
methodologies for determining core operating limits. As such, the
changes play an important role in the analysis of postulated accidents
but none of the changes affect plant hardware or the operation of plant
systems in a way that could initiate an accident. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. The proposed changes do not involve a significant reduction in
the margin of safety.
In reviewing and approving the methods used for safety analyses,
the NRC has approved the safety analysis limits which establish the
margin of safety to be maintained. Satisfaction of event-specific
acceptance criteria ensures that the approved safety analysis limits
are met and thus provides the margin of safety. The methodology being
added to the TS demonstrates, in a conservative manner, that the event
acceptance criteria are satisfied. Therefore, including this method in
the TS does not change the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 701 South Cooper, P.O.
Box 19497, Arlington, Texas 76019.
Attorney for licensee: George L. Edgar, Esq., Newman and
Holtzinger, 1615 L Street, NW., Suite 1000, Washington, DC 20036.
NRC Project Director: William D. Beckner.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: June 9, 1994.
Description of amendment request: The proposed change would revise
the Technical Specifications (TS) for the North Anna Power Station,
Units No. 1 and No. 2 (NA-1&2). Specifically, the proposed change would
relocate the TS tables of the response time limits for the Reactor Trip
System (RTS) and the Engineered Safety Feature Actuation System (ESFAS)
to station-controlled documents.
On December 29, 1993, the NRC issued Generic Letter 93-08 titled
``Relocation of Technical Specification Tables of Instrument Response
Time Limits.'' This generic letter provides guidance for preparing a
proposed license amendment to relocate the tables of response time
limits for the RTS and the Engineered ESFAS instruments from TS to
station-controlled documents.
The RTS and the ESFAS provides the signals needed to actuate the
safety equipment necessary to mitigate accidents and transients.
Consistent with Generic Letter 93-08 the licensee is requesting license
amendments for NA-1&2 to relocate the RTS and ESFAS tables of
instrument response time limits from TS to station-controlled
documents.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Specifically, operation of North Anna Power Station in accordance
with the proposed Technical Specification[s] changes will not:
1. Involve a significant increase in the probability of occurrence
or consequences of an accident previously evaluated.
The Reactor Trip System and the Engineered Safety Features
Actuation System provide the signals needed to actuate the safety
equipment necessary to mitigate accidents and transients. The proposed
changes relocate the RTS and ESFAS instrument response time limits from
the Technical Specifications to station controlled documents but will
not change the operability or surveillance requirements for these
instruments. With these proposed changes, revisions to the response
times for these instruments can be made pursuant to 10 CFR 50.59
without Nuclear Regulatory Commission approval unless the revision
involves an unreviewed safety question.
The proposed changes will not change any accident initiators or the
consequences of any analyzed accident. Therefore, the proposed changes
do not involve a significant increase in the probability or
consequences of an accident previously analyzed.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes relocate the RTS and ESFAS instrument response
time limits from the Technical Specifications to station controlled
documents but will not change the functions of these instruments. The
proposed change does not represent a change in the configuration or
operation of the plant. No new hardware is being added to the plant as
part of the proposed changes. The Technical Specifications will
continue to require the same operability and surveillance requirements
to be met for these instruments. Therefore, the proposed changes do not
create the possibility of a new or different type of accident from any
accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes will not affect the functions of the RTS or
the ESFAS instruments. Relocating the response time limits will not
alter the operability or the surveillance requirements of these
instruments. The administrative change control provisions for plant
procedures written pursuant to 10 CFR 50.59 are adequate to control
revisions to the response time limits. Therefore, the proposed changes
do not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219.
NRC Project Director: Herbert N. Berkow.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: June 9, 1994.
Description of amendment request: The proposed changes to the
Technical Specifications include: (1) modification of the high head
charging pumps seal cooling subsystem, (2) restructuring of the
Chemical and Volume Control System and Safety Injection System
Specifications, (3) relocation of certain specification requirements
within existing specifications, (4) specification of a minimum boric
acid solution temperature in lieu of heat tracing channel operability,
and (5) minor wording changes which are administrative in nature for
consistency in terminology, capitalization of defined terms and
clarification.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Specifically, operation of Surry Power Station in accordance with
the proposed Technical Specifications changes will not:
1. Involve a significant increase in the probability of occurrence
or consequences of an accident previously evaluated.
The modifications to the charging pumps and elimination of the
charging pump component cooling subsystem do not increase the
probability of occurrence of any accident or malfunction previously
evaluated in the safety analysis report. The charging pump
modifications utilize a passively designed process flow cooling
arrangement to reduce exposure, improve reliability, and improve
operability. The charging pump modifications will not decrease the
pumps ability or the associated subsystems ability to perform their
design function.
The restructuring of the Chemical and Volume Control System and
Safety Injection System specifications on a subsystem basis continues
to ensure that the reactor can be made subcritical from any operating
condition and provide sufficient shutdown margin to preclude
inadvertent criticality when in the shutdown condition. The Safety
Injection System subsystems continue to maintain sufficient boration
capability to mitigate reactivity transients within the design limits
associated with postulated accident conditions. The Safety Injection
System subsystems ensure that sufficient emergency core cooling
capability will be available in the event of a LOCA [loss-of-coolant
accident] assuming the loss of one subsystem through any single failure
consideration. Either subsystem operating in conjunction with the
accumulators remains capable of supplying sufficient core cooling to
limit the peak cladding temperatures within acceptable limits in
accordance with the loss-of-coolant accident analyses.
The Chemical and Volume Control System remains capable of achieving
Cold Shutdown of both units during any operating conditions in
accordance with the safety analysis with a minimum specified solution
temperature of 112 degrees F. Heat tracing is not required for
operability of the Safety Injection System nor does it affect the
ability of the Safety Injection System to mitigate the consequences of
any postulated accident identified in the safety analysis.
The changes ensure that the refueling water storage tank remains
capable of providing a sufficient supply of borated water for injection
by the emergency core cooling system in the event of a LOCA. The limits
specified for refueling water storage tank volume and boron
concentration continue to ensure that sufficient solution is available
within containment for recirculation cooling flow to the core, and that
the reactor will remain subcritical in Cold Shutdown consistent with
the LOCA analyses.
The specified allowed outage time of 72 hours for an inoperable
Chemical and Volume Control System subsystem or Safety Injection System
subsystem is reasonable for the repair of affected components and is
consistent with NRC Memorandum, ``Recommended Interim Revisions to
LCO's for ECCS Components,'' dated December 1, 1975, and NUREG-1431,
Standard Technical Specifications for Westinghouse Pressurized Water
Reactors. A reliability analysis (reference NRC memo above) has shown
that the impact of having one subsystem inoperable is sufficiently
small to justify continued operation for 72 hours. Engineering
evaluation of the proposed changes determined that they are bounded by
existing safety analyses. Furthermore, the proposed changes do not
increase the allowed outage times to achieve Cold Shutdown presently
specified in the Technical Specifications.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The passively designed once-through process flow cooling
arrangement for the charging pumps' seals are recommended by the pump
manufacturer with the pump seal manufacturer's concurrence and result
in no decrease in the pumps ability to perform their safety function.
Our Engineering evaluation has determined that the affected systems
ability to mitigate the consequences of any accident as described in
the safety analyses is not reduced. The restructuring and relocation of
specifications has not reduced any limiting condition for operation or
surveillance specification requirements. Changes in allowed outage
times are consistent with NUREG-0452, NUREG-1431, or Generic Letter 93-
05 and our accident analyses. Consequently, the possibility of a new or
different kind of accident is not created.
3. Involve a significant reduction in a margin of safety.
The modifications to the charging pumps result in improved
reliability and improved operability of the CVCS [chemical and volume
control system] and SI [safety injection] subsystems. Our Engineering
evaluation of the manufacturer's proposed modification and the pump
seal manufacturer's concurrence with the modification, have determined
this modification to be acceptable with no reduction in the pump's
safety-related function. The charging pump modification does not reduce
the margin of safety in any part of Technical Specifications or the
accident analyses.
The restructuring of the Chemical and Volume Control System
specifications continue to ensure that the reactor can be made
subcritical from any operating condition and provide sufficient
shutdown margin to preclude inadvertent criticality when in the
shutdown condition. The Chemical Volume and Control System remains
capable of achieving Cold Shutdown of both units during any operating
conditions in accordance with the safety analysis with a minimum
specified solution temperature of 112 degrees F. The revised allowed
outage times for the Safety Injection System subsystems do not impact
the margin of safety * * * in the Technical Specifications bases or the
accident analyses.
The Safety Injection System subsystems continue to maintain
sufficient boration capability to mitigate reactivity transients within
the design limits associated with postulated accident conditions
described within the safety analysis report. The Safety Injection
System subsystems ensure that sufficient emergency core cooling
capability will be available in the event of a LOCA assuming the loss
of one subsystem through any single failure consideration. Either
subsystem operating in conjunction with the accumulators remains
capable of supplying sufficient core cooling to limit the peak cladding
temperatures within acceptable limits in accordance with the loss-of-
coolant accident analyses.
The changes ensure that the refueling water storage tank remains
capable of providing a sufficient supply of borated water for injection
by the emergency core cooling system in the event of a LOCA. The limits
specified for refueling water storage tank volume and boron
concentration continue to ensure that sufficient solution is available
within containment for recirculation cooling flow to the core, and that
the reactor will remain subcritical in Cold Shutdown consistent with
the LOCA analyses. Consequently, the proposed change to Technical
Specifications does not involve a significant reduction [in] the margin
of safety within the accident analyses.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185.
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219.
NRC Project Director: Herbert N. Berkow.
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of amendment request: May 17, 1994.
Description of amendment request: The proposed amendment would
revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS)
3.3.c by separating the Internal Containment Spray (ICS) and Spray
Additive Systems into two distinct specifications. The proposed
amendment would also remove the requirement that for a spray train to
be operable, a spray pump suction flow path from the additive tank is
needed. In addition, the allowable out of service time for the Spray
Additive System would be increased from 48 hours to 72 hours.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
The proposed changes were reviewed in accordance with the
provisions of 10 CFR 50.92 to show no significant hazards exist. The
proposed changes will not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The likelihood that an accident will occur is neither increased nor
decreased by these TS changes. These TS changes will not impact the
function or method of operation of plant equipment. Thus, there is not
a significant increase in the probability of a previously analyzed
accident due to these changes. No systems, equipment, or components are
affected by the proposed changes. Thus, the consequences of the
malfunction of equipment important to safety previously evaluated in
the Updated Safety Analysis Report (USAR) are not increased by these
changes.
The proposed changes have no impact on accident initiators or plant
equipment, and thus, do not affect the probabilities or consequences of
an accident.
(2) Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed TS changes would not create the possibility of a new
or different kind of accident from any accident previously evaluated.
The proposed changes do not involve changes to the physical plant
or operations. Since these changes do not contribute to accident
initiation, they do not produce a new accident scenario or produce a
new type of equipment malfunction. Also, these changes do not alter any
existing accident scenarios; they do not affect equipment or its
operation, and thus, do not create the possibility of a new or
different kind of accident.
(3) Involve a significant reduction in the margin of safety.
Operation of the facility in accordance with the proposed TS would
not involve a significant reduction in a margin of safety. The proposed
changes do not affect plant equipment or operation. Safety limits and
limiting safety system settings are not affected by these proposed
changes. Extending the time the Spray Additive System may be out of
service from 48 hours to 72 hours and removing the requirement to have
a spray pump suction flow path from the additive tank for a spray train
to be operable is consistent with STS. The STS only require that the
spray system be capable of taking suction from the refueling water
storage tank and the containment sump.
The Containment Spray System would still be available and would
remove some iodine from the containment atmosphere in the event of a
Design Basis Accident. The 72 hour completion time takes into account
the Containment Spray System redundant flow path capabilities and the
low probability of the worst case Design Basis Accident occurring
during this period.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Wisconsin
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin
54301.
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P. O. Box 1497, Madison, Wisconsin 53701-1497.
NRC Project Director: John N. Hannon.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of amendment request: March 29, 1994.
Description of amendment request: The proposed amendment would
modify Point Beach Nuclear Plant Technical Specification (TS) 15.3.2,
``Chemical and Volume Control System,'' by eliminating the necessity
for high concentration boric acid and removing the operability
requirements for the associated heat tracing. The basis for Section
15.3.2 and applicable surveillances in Table 15.4.1-2 would also be
revised to support the above changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Operation of this facility under the proposed Technical
Specifications change will not create a significant increase in the
probability or consequences of an accident previously evaluated.
Reduced boron concentration in the boric acid storage tank (BAST)
is offset by increasing the volume of boric acid solution that must be
contained in the tanks. The heat tracing requirements are used to
ensure that the dissolved boric acid is maintained in solution and
available for injection into the RCS to adjust core reactivity
throughout core life and to meet GDC requirements. Chemical analyses of
boron concentrations of 4.0 weight percent have shown that
the boric acid does not crystallize at temperatures above 57 deg.F.
Ambient temperatures in the areas of the primary auxiliary building
where these components are located will normally remain above this
temperature. Hence, heat tracing will no longer be needed for boric
acid concentrations with corresponding solubility temperatures less
than ambient temperatures. The proposed Technical Specifications
requirements for boron concentration, volume, and temperature of the
BASTs and boration paths ensure that the capability to inject boric
acid is maintained. Since the components (and their function) necessary
to achieve a safe shutdown have not been changed or modified, this
change does not significantly increase the probability or consequences
of any accident previously evaluated.
The proposed changes to the boric acid system Technical
Specifications requirements for the chemical and volume control system
(CVCS) do not affect the requirements for the emergency core cooling
system (ECCS). The original design of the high head safety injection
(SI) system used the BASTs as its initial suction source. Westinghouse
WCAP-12602, ``Report For The Reduction of SI System Boron
Concentration,'' and a 10 CFR 50.59 Safety Evaluation Report performed
by Wisconsin Electric justifies the design change to use the refueling
water storage tank (RWST) as the initial suction source of SI fluid
rather than the BAST. The affected FSAR Chapter 14 accident analyses
include the Loss of Coolant Accident (LOCA) events and the Steamline
Break (SLB) events. The LOCA events are affected with respect to the
large-break post-LOCA long-term core cooling subcriticality
requirement. The SLB events are affected with respect to core
integrity. The events were analyzed assuming the elimination of the
logic which automatically opened the valves in the flow path from the
BASTs to the SI pumps upon the receipt of a safety injection signal.
The results show that we remain within the acceptance criteria of the
aforementioned FSAR Chapter 14 accident analyses. Therefore, the
proposed changes will not create a significant increase in the
probability or consequences of a[n] accident previously evaluated.
2. Operation of this facility under the proposed Technical
Specifications change will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The reactivity control function of the boron in the CVCS and SI
systems is not being changed. Therefore, the proposed changes will not
create the possibility of a new or different kind of accident from any
accident previously evaluated.
3. Operation of this facility under the proposed Technical
Specifications change will not create a significant reduction in a
margin of safety.
The intent of the proposed Technical Specifications is to ensure
that two independent flow paths from the borated water source(s) (BASTs
and/or RWST) to the reactor coolant system are maintained whenever a
unit is taken critical. This requires that sufficient quantities of
boric acid be stored in the tanks, and that this borated water can be
delivered to the reactor coolant system when required. Although we
presently require diverse sources of borated water (BAST and RWST), the
proposed reduction in diversity will be offset by the significant
increase in reliability of the boric acid system due to operation with
lower boric acid concentrations and, hence, a much lower probability of
boron precipitation and system ``freeze-up.'' Reducing the boric acid
concentration in the BASTs has been compensated for by increasing the
required volume of boric acid.
The proposed Technical Specifications requirements for boric acid
concentration and volume include the additional specification of
minimum temperature that must be maintained to assure boric acid
solubility. The minimum temperature requirement is more appropriate
than the requirement for heat tracing because it is a more precise
means of verifying and assuring solubility. Therefore, the proposed
boric acid concentration table, which includes the volume and
temperature requirements, is an appropriate substitute for the heat
tracing requirements. Although the heat tracing requirement is being
eliminated, the boric acid heat tracing system will be available during
our transition to the lower boric acid concentration to assist in
maintaining boric acid system temperature if necessary. Since our
analyses have shown that the existing FSAR Chapter 14 accident analyses
remain bounded under the proposed specifications, the margin of safety
for the plant is not significantly reduced.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John N. Hannon.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of amendment request: May 26, 1994.
Description of amendment request: Point Beach Nuclear Plant is
installing two additional emergency diesel generators and reconfigure
portions of the 4160 Volt emergency electrical power system. The
proposed amendment would revise the Point Beach Nuclear Plant Technical
Specifications (TS) to establish the requirements for the electrical
systems at Point Beach such that the TS will provide the appropriate
guidance for all interim configurations and the final configuration.
The majority of changes are incorporated in TS Section 15.3.7,
``Auxiliary Electrical Systems.'' Other Sections modified are 15.3.0,
``General Considerations,'' 15.3.14, ``Fire Protection System,'' and
15.4.6, ``Emergency Power System Periodic Tests.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
(1) Operation of this facility under the proposed Technical
Specifications will not create a significant increase in the
probability or consequences of an accident previously evaluated.
The Point Beach Nuclear Plant Final Safety Analysis Report (PBNP
FSAR) shows that the original emergency diesel generators and the
associated support systems and connections do not cause or affect the
probability of any accident evaluated in the PBNP FSAR. The additional
emergency diesel generators, the associated support systems and
connections, and reconfiguration of the emergency AC power system will
not change this. The emergency AC power system does not initiate any
accident previously evaluated in the PBNP FSAR.
The limiting conditions for operation and allowable outage times
proposed in this license amendment request are consistent with the
current requirements in the PBNP Technical Specifications. The proposed
change in the required emergency diesel generator (EDG) inspection
interval, from annually to the time as recommended by the EDG
manufacturer, will continue to maintain the operability and reliability
of the EDGs. Therefore, the probability of occurrence of an accident
previously evaluated in the FSAR is not increased by the proposed
Technical Specifications.
The consequences of the accidents previously evaluated in the PBNP
FSAR are determined by the results of analyses that are based on
initial conditions of the plant, the type of accident, transient
response of the plant, and the operation and failure of equipment and
systems. The new emergency diesel generator installation will meet the
requirements for emergency power sources for PBNP.
General Design Criterion (GDC) 39 as described in the PBNP FSAR,
states, ``An emergency power source shall be provided and designed with
adequate independency, redundancy, capacity, and testability to permit
the functioning of the engineered safety features and protection
systems required to avoid undue risk to the health and safety of the
public. This power source shall provide this capacity assuming a
failure of a single component.''
The limiting conditions for operation and allowable outage times
proposed in this license amendment request are consistent with the
requirements in GDC-39 and the current Technical Specifications for
PBNP. Therefore, this proposed license amendment does not affect the
consequences of any accident previously evaluated in the PBNP FSAR,
because the factors that are used to determine the consequences of
accidents are not being changed.
(2) Operation of this facility under the proposed Technical
Specifications change will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The PBNP FSAR shows that the original emergency diesel generators
and the associated support systems and connections do not cause any
accident evaluated in the PBNP FSAR. The additional emergency diesel
generators, the associated support systems and connections, and
reconfiguration of the emergency AC power system will not change this,
because the new emergency diesel generators will meet the requirements
for emergency power sources for PBNP. Additionally, these changes do
not introduce any type of system or component malfunction that would
create the possibility of a new or different kind of accident from any
accident previously evaluated.
The limiting conditions for operation and allowable outage times
proposed in this license amendment request are consistent with the
requirements in GDC-39 and the current Technical Specifications for
PBNP. The proposed change in the required EDG inspection interval, from
annually to the time as recommended by the EDG manufacturer, will
continue to maintain the operability and reliability of the EDGs.
Therefore, the proposed Technical Specification changes for the
addition of two diesel generators and changing the required EDG
inspection interval do not create the possibility of an accident of a
different type than any previously evaluated in the FSAR.
(3) Operation of this facility under the proposed Technical
Specifications change will not create a significant reduction in a
margin of safety.
The new diesel generator and emergency AC power system
reconfiguration design and installation are being and have been
performed to meet or exceed the original system design requirements.
The emergency diesel generators provide power to the safety equipment
that operates to maintain the margins of safety. The new diesel
generators and emergency AC power configuration will continue to
satisfy this requirement.
The limiting conditions for operation and allowable outage times
proposed in this license amendment request are consistent with the
requirements in GDC-39 and the current Technical Specifications for
PBNP. The proposed change in the required EDG inspection interval, from
annually to the time as recommended by the EDG manufacturer, will
continue to maintain the operability and reliability of the EDGs.
Therefore, the proposed Technical Specification changes for the
addition of two diesel generators and changing the required EDG
inspection interval do not reduce the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John N. Hannon.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two Creeks,
Manitowoc County, Wisconsin
Date of amendment request: May 26, 1994.
Description of amendment request: The proposed amendment would
revise the Point Beach Nuclear Plant Technical Specifications (TSs) by
extending the operation of both units with the current heatup and
cooldown limit curves to 23.6 effective full power years (EFPY). The
proposal also would revise the bases for TS Section 15.3.1.B,
``Pressure/Temperature Limits,'' to reflect the methodology for the
curve compilation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
We have evaluated these proposed amendments in accordance with the
requirements of 10 CFR 50.91(a), against the standards of 10 CFR 50.92,
and have determined that these modifications will not result in a
significant hazards consideration. A proposed amendment will not
involve a significant hazards consideration if it does not (1) involve
a significant increase in the probability or consequences of an
accident previously evaluated, (2) create the possibility of a new or
different kind of accident from any accident previously evaluated, or
(3) involve a significant reduction in a margin of safety.
The proposed heatup and cooldown curves are identical to the
current heatup and cooldown curves except for their projected
expiration. The curves were calculated using the most limiting weld and
fluence information from either unit as input to the acceptable
methodology of Regulatory Guide 1.99, Revision 2. The consequences or
probability of a previously evaluated accident will, therefore, not
significantly be increased or a margin of safety reduced.
The underlying purpose of these curves is to define an acceptable
operating range of pressures and temperatures to protect the reactor
vessels against non-ductile failure. Since this purpose remains
unchanged, a new or different kind of accident cannot be created.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John N. Hannon.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Illinois Power Company and Soyland Power Cooperative, Inc., Docket No.
50-461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois
Date of amendment request: June 20, 1994.
Brief description of amendment request: The proposed amendment
would modify Technical Specification 3/4.4.3.1, ``Reactor Coolant
Leakage--Leakage Detection Systems,'' to permit continued plant
operation with inoperable drywell floor drain sump flow rate monitoring
instrumentation. Continued plant operation would be permitted until the
first time the plant is required to be brought to COLD SHUTDOWN after
July 10, 1994.
Date of publication of individual notice in Federal Register: June
22, 1994 (59 FR 32247).
Expiration date of individual notice: July 22, 1994.
Local Public Document Room location: Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC 20555, and at the local public document
rooms for the particular facilities involved.
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of application for amendment: October 19, 1993.
Brief description of amendment: This amendment removes the scram
and Group I isolation valve closure functions associated with the main
steamline radiation monitors.
Date of issuance: June 21, 1994.
Effective date: June 21, 1994.
Amendment No.: 154.
Facility Operating License No. DPR-35: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 24, 1993 (58
FR 62151).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 21, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will
County, Illinois
Date of application for amendments: March 7, 1994, as superseded by
your submittal dated March 24, 1994.
Brief description of amendments: The amendments change Technical
Specification (TS) 4.6.1.2, ``Containment Leakage,'' by removing the
specific requirement that containment Type A leak testing be performed
at 40 10 month intervals. The revised TS now references
Appendix J to 10 CFR 50 as governing the performance of Type A testing.
Date of issuance: June 30, 1994.
Effective date: June 30, 1994.
Amendment Nos.: 62, 62, 52, and 52.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: April 28, 1994 (59 FR
22002).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 30, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: For Byron, the Byron Public
Library, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for
Braidwood, the Wilmington Township Public Library, 201 S. Kankakee
Street, Wilmington, Illinois 60481.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois Docket
Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and
2, Rock Island County, Illinois
Date of application for amendments: March 11, 1994.
Brief description of amendments: The proposed amendments revise
Technical Specification 3/4.7.D, ``Primary Containment Isolation
Valves'' by adding check valves installed in the reference leg
instrumentation line to the Limiting Condition for Operation (LCO)
statement of the Technical Specifications. The valves have been
installed as part of the modifications required to meet NRC Bulletin
93-03, ``Resolution of Issues Related to Reactor Vessel Water Level
Instrumentation in BWRs,'' dated May 28, 1993.
Date of issuance: July 6, 1994.
Effective date: For Dresden, Unit 2: the license amendment is
effective as of the date of its issuance; to be implemented when the
modifications are complete and prior to restart from any cold shutdown
after June 30, 1994, or restart from the 14th refuel outage, which ever
is first. For Dresden, Unit 3: the license amendment is effective as of
the date of its issuance; to be implemented within 30 days. For Quad
Cities, Unit 1: the license amendment is effective as of the date of
its issuance; to be implemented within 30 days. For Quad Cities, Unit
2: the license amendment is effective as of the date of its issuance;
to be implemented prior to restart following the 13th refueling outage.
Amendment Nos.: For Dresden, Unit 2: Amendment No. 128; for
Dresden, Unit 3: Amendment No. 122; for Quad Cities, Unit 1: Amendment
No. 148; for Quad Cities, Unit 2: Amendment No. 144.
Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30.
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: April 13, 1994 (59 FR
17593).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 6, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: For Dresden, the Morris Public
Library, 604 Liberty Street, Morris, Illinois 60450; for Quad Cities,
the Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 61021.
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck
Plant, Middlesex County, Connecticut
Date of application for amendment: January 28, 1994.
Brief description of amendment: The amendment modifies Surveillance
Requirement 4.6.1.2.d, regarding the Appendix J testing requirements
for the purge supply and exhaust valves, and removes surveillance
requirement 4.6.1.2.f regarding the purge and exhaust valves.
Date of issuance: June 27, 1994.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 173.
Facility Operating License No. DPR-61. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 25, 1994 (59 FR
27051).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated June 27, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Russell Library, 123 Broad
Street, Middletown, Connecticut 06457.
Consolidated Edison Company of New York, Docket No. 50-003 and Docket
No. 50-247, Indian Point Nuclear Generating Unit Nos. 1 and 2,
Westchester County, New York
Date of application for amendments: September 29, 1993.
Brief description of amendments: The amendments revise the
Technical Specifications (TSs) Administrative sections dealing with the
administrative control of keys for doors preventing unauthorized access
to High Radiation Areas in which the intensity of radiation exceeds
1000 mrem/hr. Specifically, the amendments revise TS Section 4.1.8.1.b
for Indian Point Generating Unit No. 1 and TS Section 6.12.1.b for
Indian Point Generating Unit No. 2 to add the Radiation Protection
Manager as one of the two positions which can administratively control
the keys.
Date of issuance: July 7, 1994.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 43 and 171.
Facility Operating License Nos. DPR-5 and DRP-26: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 24, 1993 (58
FR 62153).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 7, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County,
Michigan
Date of application for amendment: March 29, 1994, as corrected
April 26, 1994.
Brief description of amendment: The amendment revises the
surveillance requirements for scram discharge volume vent and drain
valves and isolation actuation instrumentation and modifies the
required actions and surveillance requirements for the emergency diesel
generators to reduce testing during power operation. These changes are
in accordance with guidance contained in Generic Letter (GL) 93-05,
``Line-Item Technical Specifications Improvements to Reduce
Surveillance Requirements for Testing During Power Operation,'' dated
September 27, 1993.
Date of issuance: June 28, 1994.
Effective date: June 28, 1994, with full implementation within 45
days.
Amendment No.: 99.
Facility Operating License No. NPF-43. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: May 25, 1994 (59 FR
27053).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 28, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161.
Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County,
Michigan
Date of application for amendment: April 26, 1994.
Brief description of amendment: The amendment revises the limiting
conditions for operation and surveillance requirements in the Technical
Specifications (TS) to delete reference to instrument response time
limit tables for the reactor protection system, instrument actuation
system and emergency core cooling system. These tables are also being
moved from the TS to the updated final safety analysis report.
Date of issuance: June 29, 1994.
Effective date: June 29, 1994, with full implementation within 60
days.
Amendment No.: 100.
Facility Operating License No. NPF-43. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: May 25, 1994 (59 FR
27053).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 29, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161.
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: March 30, 1994.
Brief description of amendments: The amendments revise Technical
Specification Table 4.3-3 to allow the analog channel operational test
interval for radiation monitoring instrumentation to be increased from
monthly to quarterly and are consistent with the guidance in Generic
Letter 93-05, ``Line-Item Technical Specifications Improvements to
Reduce Surveillance Requirements for Testing During Power Operation.''
Date of issuance: July 5, 1994.
Effective date: July 5, 1994.
Amendment Nos.: 121/115.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 25, 1994 (59 FR
27054).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 5, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant Units 3 and 4, Dade County, Florida
Date of application for amendments: April 19, 1994.
Brief description of amendments: These amendments change the
surveillance interval specified for air or smoke flow test through the
containment spray header from once per 5 years to once per 10 years.
Date of issuance: June 28, 1994.
Effective date: June 28, 1994.
Amendment Nos. 165 and 159.
Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 25, 1994 (59 FR
27055).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 28, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant Units 3 and 4, Dade County, Florida
Date of application for amendments: November 25, 1992 as
supplemented by letter dated March 4, 1994.
Brief description of amendments: These amendments implement Generic
Letter 90-06, ``Resolution of Generic Issue 70, `Power-Operated Relief
Valve and Block Valve Reliability,' and Generic Issue 94, `Additional
Low-Temperature Overpressure Protection for Light-Water Reactors,'
Pursuant to 10 CFR 50.54(f).''
Date of issuance: June 28, 1994.
Effective date: June 28, 1994.
Amendment Nos. 166 and 160.
Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 14, 1993 (58 FR
19478).
The licensee's letter of March 4, 1994 did not change the no
significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 28, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199.
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile Island
Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of application for amendment: June 12, 1991.
Brief description of amendment: The amendment revises the plant
Technical Specifications to clarify the setpoint ranges for the
pressurizer power-operated relief valve and provides action statements
to be satisfied when setpoint ranges are not met.
Date of issuance: June 30, 1994.
Effective date: As of its date of issuance to be implemented within
60 days.
Amendment No.: 186.
Facility Operating License No. DPR-50. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 10, 1991 (56 FR
31435).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 30, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, Walnut Street and Commonwealth
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.
Illinois Power Company and Soyland Power Cooperative, Inc., Docket No.
50-461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois
Date of application for amendment: June 20, 1994.
Brief description of amendment: The amendment revised the Technical
Specifications by modifying a footnote to Technical Specification 3/
4.4.3.1, ``Reactor Coolant System Leakage--Leakage Detection Systems,''
to permit continued plant operations with inoperable drywell floor
drain sump flow monitoring instrumentation until the first time the
plant is required to be brought to cold shutdown after July 10, 1994.
Date of issuance: July 8, 1994.
Effective date: July 8, 1994.
Amendment No.: 90.
Facility Operating License No. NPF-62. The amendment revised the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: Yes (59 FR 32247 dated June 22, 1994). That notice
provided an opportunity to submit comments on the Commission's proposed
no significant hazards consideration determination. No comments have
been received. The notice also provided for an opportunity to request a
hearing by July 22, 1994, but indicated that if the Commission makes a
final no significant hazards consideration determination any such
hearing would take place after issuance of the amendment.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, and final determination of no significant
hazards consideration are contained in a Safety Evaluation dated July
8, 1994.
Local Public Document Room location: The Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727.
Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: February 10, 1992, as
supplemented April 14, 1994.
Brief description of amendment: The amendment removes two tables
from the Technical Specifications which list reactor trip system
instrumentation response times and engineered safety features actuation
system instrumentation response times. These tables will be placed in
the Millstone 3 Technical Requirements Manual.
Date of issuance: June 28, 1994.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 91.
Facility Operating License No. NPF-49. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 25, 1994 (59 FR
27058).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 28, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, Connecticut 06360.
Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: March 23, 1994.
Brief description of amendment: The amendment modifies Technical
Specification Table 3.7-6, ``Area Temperature Monitoring,'' by creating
two zones for the main steam valve building (MSVB) and increasing the
maximum normal excursion temperature limit for this area from 120 deg.F
to 140 deg.F. Technical Specification Table 3.7-6 currently identifies
the entire MSVB with a temperature limit of 120 deg.F.
Date of issuance: June 29, 1994.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 92.
Facility Operating License No. NPF-49. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 28, 1994 (59 FR
22009).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 29, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Community-Technical College, Thames Valley Campus, 574 New London
Turnpike, Norwich, Connecticut 06360.
Northern States Power Company, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of application for amendment: January 4, 1994, as supplemented
March 28, 1994.
Brief description of amendment: The amendment modifies Section
3.11, Reactor Fuel Assemblies, by removing information concerning the
analytical method to determine average planar linear heat generation
rate and adding a reference to the Core Operating Limits Report. In
Section 6.7, Reporting Requirements, the listing of approved analytical
methods for developing the Core Operating Limits Report is revised and
the specific version of the analytical methods used to develop the
report is identified. Also, the Bases for Section 3.11 concerning the
calculational methodology for minimum critical power ratio was revised.
Date of issuance: June 30, 1994.
Effective date: June 30, 1994.
Amendment No.: 88.
Facility Operating License No. DPR-22. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 2, 1994 (59 FR
10011).
The March 28, 1994, letter provided clarifying information that was
within the scope of the March 2, 1994, notice. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
June 30, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power
Plant, Unit 3, Humboldt County, California
Date of application for amendment: July 7, 1993. (Reference HBL-93-
041)
Brief description of amendment: This amendment modified the
Technical Specifications incorporated in Facility Operating License No.
DPR-7 as Appendix A by revising technical specification VII.H.3,
``Semiannual Radioactive Effluent Release Report,'' to extend the
reporting period from semiannually to annually and to change the report
submission date from 60 days after January 1 and July 1 of each year to
before April 1 of each year.
Date of issuance: June 30, 1994.
Effective date: This license amendment is effective as of the date
of its issuance and must be fully implemented no later than 30 days
from the date of issuance.
Amendment No.: 26.
Facility Operating License No. DPR-7: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 19, 1994 (59 FR
2869).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 30, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Humboldt County Library, 636 F
Street, Eureka, California 95501.
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: September 1, 1992 and October
15, 1992 as supplemented by letters dated October 30, 1992, March 16,
1993, June 10, 1993, July 28, 1993, September 10, 1993, April 29, 1994,
June 2, 1994, June 9, 1994, and June 15, 1994.
Brief description of amendments: The amendments extend the interval
for certain Technical Specifications surveillance requirements to 24
months with an additional 25-percent grace period.
Date of issuance: June 28, 1994.
Effective date: As of 30 days after the date of issuance.
Amendment Nos.: 71 and 34.
Facility Operating License Nos. NPF-39 and NPF-85: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 16, 1992 (57
FR 42778) and October 28, 1992 (57 FR 48823).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 28, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Philadelphia Electric Company, Docket No. 50-352, Limerick Generating
Station, Unit 1, Montgomery County, Pennsylvania
Date of application for amendment: May 6, 1994, as supplemented by
letter dated June 3, 1994.
Brief description of amendment: This amendment revised TS Section
5.5.3, ``Capacity,'' to facilitate an interim increase in the Unit 1
Spent Fuel Pool from 2040 fuel assemblies to 2500 fuel assemblies.
Date of issuance: June 30, 1994.
Effective date: June 30, 1994.
Amendment No.: 72.
Facility Operating License No. NPF-39. This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 25, 1994 (59 FR
27063).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 30, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: December 22, 1993.
Brief description of amendment: The amendment removes the reference
to American Society for Testing and Materials (ASTM) Standard D 270-65
from Technical Specification Surveillance Requirement 4.12A.1.i. ASTM D
270-65, which specifies procedures to draw a representative fuel oil
sample, has been superseded and is no longer in effect. The amendment
will allow ASTM D 4057-88 or subsequent industry standards to be used
for the sampling of diesel fire pump fuel oil.
Date of issuance: June 27, 1994.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 214.
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 2, 1994 (59 FR
4944).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 27, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: January 31, 1994.
Brief description of amendment: The amendment revises Technical
Specification (TS) 3.8, ``Miscellaneous Radioactive Materials
Sources,'' to adopt the Limiting Conditions for Operation of Section 3/
4.7.6, ``Sealed Source Contamination,'' in NUREG-0123, ``Standard
Technical Specifications for General Electric Boiling Water Reactors
(BWR/5).'' The amendment also reformats TSs 3.8 and 4.8 to make them
consistent with the remainder of the TSs.
Date of issuance: June 27, 1994.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 215.
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 2, 1994 (59 FR
10014).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 27, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County,
Alabama
Dates of application for amendments: July 19, 1993 (TS-334).
Brief description of amendments: The amendments remove the
Technical Specifications (TS) addressing reactor coolant chemistry
limits and associated sampling requirements that are applicable when
the reactor is defueled. The TS requirements being removed have been
conservatively incorporated into the BFN chemistry program as elements
of a licensee-controlled procedure. Any future changes to these
chemistry requirements must be evaluated in accordance with 10 CFR
50.59 to determine whether the changes involve an unreviewed safety
question. A change involving an unreviewed safety question would
require a license amendment and NRC review and approval prior to
implementation. In addition, changes to the reactor coolant chemistry
TS, applicable when fuel is in the reactor, are included in these
amendments to provide clarification and to ensure consistency in
requirements among units.
Date of issuance: June 28, 1994.
Effective date: June 28, 1994.
Amendment Nos.: 208, 224 and 181.
Facility Operating License Nos. DPR-33, DPR-52, and DPR-68:
Amendments revised the Technical Specifications.
Dates of initial notice in Federal Register: November 10, 1993 (58
FR 59756).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 28, 1994.
No significant hazards consideration comments received: None
Local Public Document Room location: Athens Public Library, South
Street, Athens, Alabama 35611.
Tennessee Valley Authority, Docket Nos. 50-259 and 50-296, Browns Ferry
Nuclear Plant, Units 1 and 3, Limestone County, Alabama
Date of application for amendment: July 2, 1992 (TS 314)
Brief description of amendments: The amendments revise requirements
associated with Residual Heat Removal valve pressure switches in the
Browns Ferry Units 1 and 3 Technical Specifications.
Date of issuance: June 30, 1994.
Effective date: June 30, 1994.
Amendment Nos.: 209 and 182.
Facility Operating License Nos. DPR-33 and DPR-68: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 28, 1992 (57 FR
48826).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 30, 1994.
No significant hazards consideration comments received: None.
Local Public Document Room location: Athens Public Library, South
Street, Athens, Alabama 35611.
The Cleveland Electric Illuminating Company, Centerior Service Company,
Duquesne Light Company, Ohio Edison Company, Pennsylvania Power
Company, Toledo Edison Company, Docket No. 50-440, Perry Nuclear Power
Plant, Unit No. 1, Lake County, Ohio
Date of application for amendment: February 28, 1992.
Brief description of amendment: The amendment revises Technical
Specification Table 3.3.7.1-1 and Table 4.3.7.1-1 to remove the area
criticality monitors for the fuel preparation pool, spent fuel pool,
and the upper containment pools and their associated action statements,
notes, and surveillance requirements. Editorial changes were made as
required in the tables.
Date of issuance: June 28, 1994.
Effective date: June 28, 1994.
Amendment No. 62.
Facility Operating License No. NPF-58 This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 24, 1992 (57 FR
28205).
The Commission's related evaluation of the amendment is contained
in an Environmental Assessment dated June 13, 1994, and in a Safety
Evaluation dated June 28, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, Ohio 44081.
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear
Power Station, Unit No. 1, Ottawa County, Ohio
Date of application for amendment: January 31, 1994.
Brief description of amendment: This amendment revises TS 3/4.1.1.2
to permit the reduction of boron concentration of water within the
reactor coolant system (RCS), subject to certain restrictions, when the
reactor is in Mode 5 and RCS flow is less than 2800 gpm. This amendment
is related to Amendment No. 176, which was issued by the NRC on
December 8, 1992, and incorporated a similar revision for Mode 6
operation.
Date of issuance: June 28, 1994.
Effective date: June 28, 1994.
Amendment No. 188.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 16, 1994 (59 FR
12369).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 28, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Toledo Library,
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: February 14, 1994.
Brief description of amendments: The amendments revise the
technical specifications by replacing the requirements for reporting
radiological effluents from semiannual to annual, and the report due
dates from 60 days after January 1 and July 1 to prior to May 1. The
changes are consistent with the requirements for reporting radioactive
effluent releases specified in 10 CFR 50.36a.
Date of issuance: June 1, 1994.
Effective date: June 1, 1994, to be implemented within 30 days of
issuance.
Amendment Nos.: Unit 1--Amendment No. 25; Unit 2--Amendment No.
11.
Facility Operating License Nos. NPF-87 and NPF-89. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 28, 1994 (59 FR
22016).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 1, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 701 South Cooper, P.O.
Box 19497, Arlington, Texas 76019.
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of application for amendment: February 17, 1994.
Brief description of amendment: The amendment modifies the
administrative section of the technical specifications (TS) to reflect
management and organizational changes at the Washington Public Power
Supply System (the licensee) for operation of the WNP-2 facility. The
proposed changes would (1) modify the reporting responsibility of the
quality assurance organization from the Managing Director to the
Assistant Managing Director, Operations (AMDO), and (2) modify the
appointment authority for the Corporate Nuclear Safety Review Board
(CNSRB) from the Managing Director to the AMDO. These changes are
proposed to reflect the current designation of the AMDO as the
licensee's designated official with corporate responsibility for
overall plant nuclear safety and as the direct report for the CNSRB.
In addition, the proposed change would (1) delete the specific
requirement for health physics/chemistry program procedures, (2) modify
the titles of two positions on the Plant Operations Committee (POC) to
reflect revised organizational titles, and (3) delete the requirement
that the CNSRB Executive Secretary be designated from the CNSRB
membership.
The staff denies the licensee's request to change CNSRB membership
requirements from nine personnel to a minimum of nine personnel.
Date of issuance: June 28, 1994.
Effective date: 5 days after the date of issuance.
Amendment No.: 126.
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 13, 1994 (59 FR
17609).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 28, 1994.
Public comments on proposed no significant hazards consideration
comments received: No.
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352.
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of application for amendment: May 5, 1994.
Brief description of amendment: The amendment changes equipment
numbering on three primary containment isolation valves.
Date of issuance: June 28, 1994.
Effective date: June 28, 1994.
Amendment No.: 127.
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 12, 1994 (59 FR
24762).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 28, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas.
Date of amendment request: February 24, 1994.
Brief description of amendment: The amendment revises Technical
Specification 3.9.4, Containment Building Penetrations, to allow the
use of temporary alternate closure methods for the emergency personnel
escape lock and containment wall penetrations, during core alterations
or movement of irradiated fuel within the containment.
Date of issuance: July 7, 1994.
Effective date: July 7, 1994, to be implemented within 30 days of
issuance.
Amendment No.: 74.
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 13, 1994 (59 FR
17610).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 7, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621.
Dated at Rockville, Maryland, this 13th day of July 1994.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Acting Director, Division of Reactor Projects--III/IV, Office of
Nuclear Reactor Regulation.
[FR Doc. 94-17503 Filed 7-19-94; 8:45 am]
BILLING CODE 7590-01-P
