[Federal Register Volume 62, Number 156 (Wednesday, August 13, 1997)]
[Notices]
[Pages 43365-43381]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X97-10813]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice
Applications and Amendments to Facility Operating
LicensesInvolving 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 July 19, 1997, through August 1, 1997. The
last biweekly notice was published on July 30, 1997, (62 FR 40843).
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 Chief, Rules
Review and Directives Branch, Division of Freedom of Information and
Publications Services, Office of Administration, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, 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. The filing of requests for a hearing and
petitions for leave to intervene is discussed below.
By September 12, 1997, 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 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
[[Page 43366]]
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 20555-0001, 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,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, 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, and at the local public document room for
the particular facility involved.
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: July 8, 1997
Description of amendments request: The proposed amendments remove
the suppression chamber water volume band from Technical Specification
(TS) 3.6.2.1.a.1 while retaining the equivalent water level band. The
values for the suppression chamber water volume corresponding to the
low and high suppression chamber water levels will be retained in the
Bases section of the TS and will be revised by the proposed amendments
to account for the displacement of water due to the planned
installation of larger emergency core cooling system suction strainers.
The revised relationship between the high and low suppression chamber
water levels and suppression chamber water volume will also be
described in 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 amendments do
not involve a significant increase in the probability or consequences
of an accident previously evaluated.
The proposed change revises the values of the minimum and
maximum suppression chamber pool water volume limits. The water
inventory of the suppression chamber pool is not a precursor of an
accident and, therefore, cannot increase the probability of an
accident previously evaluated. The pressure suppression chamber
water pool mitigates the consequences of loss-of-coolant accidents
(LOCAs) transients [sic], and other events by providing a heat sink
for reactor primary system energy releases. The proposed minimum and
maximum pool water volume values will be consistent with the current
suppression chamber pool water level limits. No changes to setpoints
will be made as a result of the proposed change. The impact of the
proposed change to the minimum and maximum suppression chamber pool
volume limits on the suppression chamber pool temperatures and
pressures following a design basis LOCA, an Safety/Relief Valve
(SRV) blowdown event, an Anticipated Transient Without Scram (ATWS)
event, an Appendix R fire event, and a station blackout event has
been evaluated and does not cause accident parameters to exceed
acceptable values. In addition, the impact the proposed change has
on the time to reach cold shutdown when using the alternate Residual
Heat Removal (RHR) shutdown cooling function is negligible. The
potential impact the proposed change to the suppression chamber pool
water volume limits has on SRV line loads, SRV discharge line
reflood height, wetwell pressurization, suppression chamber pool
swell loads, vent thrust loads, and condensation oscillation and
chugging loads was also reviewed. The change to the suppression
chamber pool water volume limits has no significant adverse impact
on any of these parameters. As delineated above, the capability of
the suppression chamber water pool to perform its mitigative
functions is not affected by the proposed change. Therefore, the
proposed change does not involve a significant increase in the
consequences of an accident previously evaluated.
2. The proposed amendments would not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
[[Page 43367]]
The proposed change revises the values of the minimum and
maximum volume of the suppression chamber water pool. The proposed
change will not alter any physical mechanism by which the
suppression chamber water pool volume is maintained between the
minimum and maximum values. The suppression chamber pool water level
will continue to be maintained between -27 and -31 inches. The
suppression chamber pool water level limits are retained in
Technical Specification (TS) 3.6.2.1.a.1, since this is the
information available to the operators regarding the suppression
chamber pool water volume limits. These level limits are equivalent
to the suppression chamber pool water volume limits; therefore, it
is only the presentation of the equivalency that is being relocated
to the Bases and the Updated Final Safety Analysis Report (UFSAR).
As such, the relocated suppression chamber pool water volume limits
are not required to be in the TS to provide adequate protection of
the public health and safety. As a result of the proposed strainer
changes, there are no physical changes to any other suppression
chamber components or instrumentation. No new mode of operation is
introduced as a result of the proposed change. Analyses have been
performed which conclude that the proposed change will not affect
the operability of the equipment designed to mitigate the
consequences of an accident. Therefore, the proposed change does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. The proposed license amendments do not involve a significant
reduction in a margin of safety.
The proposed change revises the values of the minimum and
maximum suppression chamber water pool volumes. The pressure
suppression chamber water pool mitigates the consequences of several
postulated accidents and transients by providing a heat sink for the
primary coolant system. These accidents and events are the
postulated design basis LOCA, an SRV blowdown event, an ATWS event,
an Appendix R fire, and station blackout events. The consequences of
the change in the suppression pool water volume limits have been
evaluated for these events, and there is no significant reduction in
the margin of safety.
The results of the analyses for the postulated accidents and
events indicate the temperature of the suppression chamber pool
water could increase slightly as a consequence of the decrease in
the minimum suppression chamber pool water volume limit. However,
the suppression chamber pool water and containment temperatures
remain within acceptable values. The impact of the calculated
increase in containment temperature on the available Net Positive
Suction Head (NPSH) for the Residual Heat Removal (RHR) and Core
Spray pumps has been evaluated for the postulated design basis LOCA
and indicate[s] adequate NPSH is maintained throughout the event.
The potential impact of the proposed change to the suppression
chamber pool water volume limits on the SRV line loads, SRV
discharge line reflood height, wetwell pressurization, suppression
chamber pool swell loads, vent thrust loads, and condensation
oscillation and chugging loads was evaluated with the conclusion
that there are no adverse impacts on these parameters.
In addition, a small suppression chamber pool water temperature
increase could result due to the reduction in minimum suppression
pool volume limit in the event reactor shutdown is conducted through
a path utilizing the suppression chamber pool. Such a shutdown path
is an alternative to the normal RHR shutdown cooling function, and
the small potential increase in temperature results in a negligible
increase in the time required to reach cold shutdown conditions.
Cold shutdown conditions can still be reached well within the
Technical Specification requirements.
The proposed increase in the suppression pool water volume limit
does not adversely impact containment parameters as a result of
postulated accidents and events. The potential increase in
temperature of the pressure suppression chamber pool water does not
significantly decrease the ability to maintain containment
parameters within acceptable limits. The potential increase in time
to reach cold shutdown conditions utilizing the suppression pool as
an alternative to the normal RHR shutdown cooling function is
negligible. Therefore, the proposed change to revise the minimum and
maximum suppression water pool volumes does not involve a
significant reduction in a margin of safety.
The suppression chamber pool water level limits are retained in
TS 3.6.2.1.a.1, since this is the information available to the
operators regarding the suppression chamber pool water volume
limits. These level limits are equivalent to the suppression chamber
pool water volume limits and the equivalency is being relocated to
the Bases and the UFSAR. As such, the relocated suppression chamber
pool water volume limits are not required to be in the TS to provide
adequate protection of the public health and 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 North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602
NRC Project Director: Gordon E. Edison, Acting
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One,
Unit No. 2, Pope County, Arkansas
Date of amendment request: July 21, 1997
Description of amendment request: Technical Specification Change
Request Concerning Emergency Feedwater Surveillance Testing. This
request is to make several changes to the ANO-2 Technical
Specifications including an extension of the emergency feedwater (EFW)
pump surveillance testing frequency, a reduction in the minimum steam
generator pressure required to perform the surveillance testing on the
turbine-driven EFW pump, and a modification to the EFW pump testing
requirements.
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 not Involve a Significant Increase in the Probability or
Consequences of an Accident Previously Evaluated.
The proposed changes included in this amendment request are
being made to the emergency feedwater (EFW) system technical
specification (TS) surveillances. These changes include surveillance
interval modifications, allowances to perform the turbine driven EFW
pump surveillance at a lower steam generator (S/G) pressure,
removing the requirements to perform specific EFW surveillance
requirements (SRs) during plant shutdowns, bases changes, and
various administrative changes. These changes are consistent with
the applicable SRs located in NUREG-1432 and have therefore, been
previously approved by the NRC.
These changes do not alter the functional characteristics of any
plant component and do not allow any new modes of operation of any
component. The accident mitigation features of the plant are not
affected by the proposed amendment request. No modifications have
been made to the EFW system due to this amendment request. Although
the minimum steam generator pressure has been reduced for the
turbine driven EFW pump testing, calculations show that significant
margin exists between the proposed value and that needed to
adequately perform the test. The capability of the EFW pumps to
perform their required safety function is not impacted by this
change. The addition of the electric driven EFW flow path
verification will help [to] assure proper alignment of both trains
of EFW following extended outages.
The accident mitigation features of the plant are not affected
by the proposed amendment. No modification has been made to the pump
or turbine driver. The capability of the turbine driven EFW pump to
perform its required function is not impacted by this change. The
EFW pumps will be tested in accordance with the more restrictive of
the
[[Page 43368]]
data points required by the safety analysis or the inservice testing
program. Therefore, this change does not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
2. Does Not Create the Possibility of a New or Different Kind of
Accident from any Previously Evaluated.
No new possibility for an accident is introduced by modifying
the proposed specifications for the surveillance testing of the EFW
pumps. The EFW surveillance requirements will continue to
demonstrated the pump's ability to perform its safety function. The
modifications to the proposed EFW surveillance requirements are
consistent with the current revision of NRC approved NUREG -1432,
``Standard Technical Specifications Combustion Engineering Plants''
(ITS). Therefore, this change does not create the possibility of a
new or different kind of accident from any previously evaluated.
3. Does Not Involve a Significant Reduction in Margin of Safety.
The safety function of the EFW system is not altered as a result
of this change. The capability of the EFW pumps to perform their
required function is not impacted by this change. The capability of
the EFW pumps is not impacted by this change. The EFW pumps will be
tested and proven operable in accordance with the more restrictive
of the data points required by the safety analysis of the inservice
testing program. The addition of the electric driven EFW flow path
verification will help assure [to] proper alignment of both trains
of EFW following extended outages. Therefore, this change does not
involve a significant reduction in the margin of safety.
Therefore, based upon the reasoning presented above and the
previous discussion of the amendment request, Entergy Operations has
determined that the requested change does not involve significant
hazards consideration.
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: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
NRC Project Director: James Clifford, Acting
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida
Date of amendment request: June 26, 1997
Description of amendment request: The proposed amendment would
revise the Operating License No. DPR-72, License Condition 2.C.(5) and
delete the requirement for installation and testing of flow indicators
in the emergency core cooling system to provide indication of 40
gallons per minute flow for boron dilution from the license. Approval
of this amendment will allow removal of the appropriate flow
indicators, DH-45-Fl and DH-46-Fl, from the Crystal River 3 (CR3) 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:
Criterion 1
The change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
This license amendment removes the requirement for flow
indication on the DH drop line and auxiliary pressurizer spray line
for boron precipitation mitigation during a LOCA [Loss of Coolant
Accident]. The original need for these indicators was to provide
flow indication to the operator to aid in decision making relative
to an alternate active method for boron precipitation prevention.
Alternate active methods have been replaced by the passive flow path
through the gaps which exist between the reactor vessel and the
reactor vessel internals. Since auxiliary pressurizer spray flow is
no longer used, and no other active means is required to be employed
by the operator in the event drop line flow is not indicated, the
original usefulness of and need for this indication no longer
exists. Removal of this requirement from the license condition does
not involve a change in the Improved Technical Specifications. The
operators do not use the flow indication for decision making in
post-accident conditions. Since these instruments are no longer used
for boron precipitation mitigation during a LOCA, abandonment or
removal of flow indicator DH-45-Fl and DH-46-Fl does not increase
the probability of an accident because no previously evaluated
accidents at CR-3 are initiated by DH-45-Fl or DH-46-Fl. Those CR-3
accidents that are analyzed are contained in the Final Safety
Analysis Report (FSAR) and include events such as Loss-of-Coolant
Accidents, Main Steam Line Breaks, Station Blackout, Anticipated
Transients Without Scram, etc. Since DH-45-Fl and DH-46-Fl are
attached to the outside of the DH drop line and auxiliary
pressurizer spray line, their removal will not change the design,
material, or construction standards applicable to the DH System
piping. The removal of the indicator will not affect overall system
performance of the ECCS. All of these previously evaluated accidents
described in the CR-3 FSAR have dose consequences which remain well
within the requirements of 10 CFR Part 100 (25 rem whole body, 300
rem thyroid) and GDC [General Design Criterion] 19 (5 rem whole
body, or its equivalent to any part of the body). Removal of DH-45-
Fl and DH-46-Fl will not alter any assumptions made in evaluating
the radiological consequences of any accident described in the FSAR
nor will it affect any fission product barriers since the ECCS and
containment systems will still perform to meet design requirements.
Therefore, removal of DH-45-Fl and DH-46-Fl will not alter the
consequences of an accident previously evaluated.
Criterion 2
The change does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed license amendment removes the requirement for
indicators which were originally installed to aid the operator in
decision making relative to an alternate flow path for boron
precipitation mitigation during a LOCA. These indicators no longer
serve this purpose, since alternate active flow paths are no longer
considered. Evaluations which consider boron precipitation no longer
rely on three active methods of mitigation, but rather one active
and one passive. Operator action is not required to effect the
backup method in the event that the primary method fails due to a
single active failure. The flow indicators are external to the DH
System piping. They do not penetrate any piping so their removal
cannot create the possibility of a new or different kind of
accident. The accident mitigation strategies remain the same
regardless of whether or not the flow indicators are present.
Therefore, the flow indicators serve no purpose in the analyses. The
proposed amendment does not affect any of the parameters or
conditions that could contribute to the initiation of any accidents.
Criterion 3
The change does not involve a significant reduction in the
margin of safety.
Boron precipitation within the reactor vessel during post-LOCA
conditions, if it were to occur, would challenge the margin of
safety that is provided by assuring compliance with Criterion 5 of
10 CFR 50.46. The license amendment does not change the methodology
of mitigating the consequences of boron precipitation following a
LOCA as described in the current licensing basis. The primary method
of flow through the DH drop line and the use of gap flow as the
``backup'' method for prevention of boron precipitation have been
analyzed, shown to meet all the criteria of 10 CFR 50.46, and
accepted by the NRC. The passive method requires no specific
operator action for initiation, in the event that the primary method
fails due to a single active failure. Therefore, the indication
serves no safety function and does 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 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: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428
[[Page 43369]]
Attorney for licensee: R. Alexander Glenn, General Counsel, Florida
Power Corporation, MAC - A5A, P. O. Box 14042, St. Petersburg, Florida
33733-4042
NRC Project Director: Frederick J. Hebdon
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida
Date of amendment request: July 18, 1997
Description of amendment request: The proposed amendment would
revise the Crystal River 3 (CR-3) technical specifications (TS) to
incorporate a new TS 3.4.11 for a Low Temperature Overpressure
Protection (LTOP) System. The proposed changes would be consistent with
the recommendations in the NRC Generic Letter 88-11, ``NRC Position on
Radiation Embrittlement of Reactor Vessel Materials and Its Impact on
Plant Operations.'' TS 3.5.3 and associated TS Bases would also be
revised to reflect the proposed change.
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 Not Involve a Significant Increase in the Probability or
Consequences of an Accident Previously Evaluated.
This change does not involve a significant increase in the
probability or consequences of any accident previously evaluated.
There are currently no LTOP requirements in the CR-3 Improved
Technical Specifications. CR-3 currently implements LTOP features
through administrative controls and a lowered PORV [power-operated
relief valve] setpoint. The proposed change will establish new LTOP
technical specification requirements necessary to preclude an LTOP
event from occurring. The proposed LTOP requirements are based on
safety analyses that apply ASME [American Society of Mechanical
Engineers] Code Case N-514. These requirements will decrease the
probability of a low temperature overpressure event by providing
protection for all pressure and temperature combinations for which a
low temperature overpressure event may be postulated.
The consequences of a low temperature overpressure accident are
not affected by this change. There is no change to the 10 CFR [Code
of Federal Regulations] Part 100 dose calculation for a low
temperature overpressure accident.
2. Does Not Create the Possibility of a New or Different Kind of
Accident from any Previously Evaluated
This change does not create the possibility of a new or
different kind of accident from any previously evaluated.
The new LTOP Technical Specification does not require
modification to the plant nor does it create a new mode of plant
operation. The LTOP system adds no new accident initiators.
3. Does Not Involve a Significant Reduction in the Margin of
Safety.
The proposed change does not involve a significant reduction in
the margin of safety and will provide added safety benefit gained
through the requirements to preclude a low temperature
overpressurization event to the RCS [reactor coolant system].
The margin of safety prior to having an LTOP system was limited
due to the informal, administrative method of minimizing the impact
of a low temperature overpressure accident. By formalizing these
requirements into a technical specification, at the least, margin of
safety is retained and perhaps improved due to the elevated
significance of required actions.
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: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428
Attorney for licensee: R. Alexander Glenn, General Counsel, Florida
Power Corporation, MAC - A5A, P. O. Box 14042, St. Petersburg, Florida
33733-4042
NRC Project Director: Frederick J. Hebdon
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida
Date of amendment request: July 29, 1997
Description of amendment request: The proposed amendment would
revise the Crystal River Nuclear Generating Unit 3 (CR3) technical
specifications (TS) to add subcooling margin and decay heat removal
(low pressure injection) flow and correct certain nomenclature in the
post-accident monitoring (PAM) instrumentation TS. In addition, the
licensee proposes to add emergency diesel generator (EDG) kilowatt (kW)
indication to the PAM instrumentation. Specifically, the following TS
would be revised:
A. Table 3.3.17-1, Function 8: The descriptor is changed from
``Containment Pressure (Narrow Range)'' to ``Containment Pressure
(Expected Post-Accident Range).''
B. Table 3.3.17-1, Function 18: The required channels for Core Exit
Temperature (Backup) is changed from ``2 sets of 5'' to ``3 per core
quadrant.''
C. Table 3.3.17-1: A new Function 20 is added and designated as
``Low Pressure Injection Flow'', with 2 required channels, and
Condition E.
D. Table 3.3.17-1: A new Function 21 is added and designated as
``Degrees of Subcooling'', with 2 required channels, and Condition E.
E. Table 3.3.17-1: A new Function 22 is added and designated as
``Emergency Diesel Generator kW Indication'', with 2 required channels,
and Condition E. A note clarifying the number of required channels is
added: ``(c): one indicator per EDG''.
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 items A, B, C, D and E
corresponds to the specific TS changes described above.
1. The proposed changes will not significantly increase the
probability or consequences of an accident previously evaluated
because:
A/B. The changes in containment pressure and core exit
thermocouple nomenclature do not reflect any physical changes to the
facility. This would have no impact on accident probability or
consequences.
C/D/E. The addition of low pressure injection flow, degrees of
subcooling, and EDG kW indication to the Post-Accident
Monitoring Instrumentation LCO [Limiting Condition for Operation] is
being done to comply with a commitment made during the technical
specification improvement program to include in the technical
specifications that instrumentation which monitors variables
classified as Type A in accordance with Regulatory Guide 1.97. These
three variables have been reclassified as Type A. The associated
instruments are used in post-accident conditions to prompt the
operators to take certain mitigative actions. Therefore, the
probability of an accident occurring is unaffected. As part of the
re-classification of these variables to Type A and inclusion in
technical specifications, the associated monitoring instrumentation
will be under more strict surveillance and control, which provides
additional assurance that the prescribed manual operator actions
will be implemented when necessary. This, in turn, assures the
previously evaluated accident consequences remain valid.
2. The proposed changes will not create the possibility of a new
or different kind of accident from any accident previously evaluated
because:
A/B. The changes in containment pressure and core exit
thermocouple nomenclature do not reflect any physical changes to
the facility. The changes provide clarification for the instruments
which are required to comply with the LCO. This would not create
possibility of a new or different kind of accident.
C/D/E.The addition of low pressure injection flow, degrees of
subcooling, and EDG kW indication to the Post-Accident Monitoring
Instrumentation LCO is being
[[Page 43370]]
done to comply with a commitment made during the technical
specification improvement program to include in the technical
specifications that instrumentation which monitors variables
classified as Type A in accordance with Regulatory Guide 1.97. These
three variables have recently been reclassified as Type A. The
associated instruments are used after an accident occurs to prompt
the operators to take certain mitigative actions. Since the
instrumentation is used only post-accident, these changes do not
create the possibility of a new or different kind of accident.
3. The proposed change will not involve a significant reduction
to the margin of safety because:
A/B. The changes in containment pressure and core exit
thermocouple nomenclature have no affect on the margin of safety.
The changes provide clarification of the technical specifications.
This reduces the potential for confusion regarding this
instrumentation.
C/D/E. The addition of low pressure injection flow, degrees of
subcooling, and EDG kW indication to the post-accident
monitoring instrumentation table in technical specifications results
in added controls on the OPERABILITY of this post-accident
monitoring instrumentation and provides greater assurance that it
will be available should an accident occur.
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: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428
Attorney for licensee: R. Alexander Glenn, General Counsel, Florida
Power Corporation, MAC - A5A, P. O. Box 14042, St. Petersburg, Florida
33733-4042
NRC Project Director: Frederick J. Hebdon
Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-
423, Millstone Nuclear Power Station, Unit No. 3, New London
County, Connecticut
Date of amendment request: July 18, 1997
Description of amendment request: The proposed amendment adds a new
Technical Specification and associated Bases to address the operability
of the steam generator atmospheric relief bypass valves (SGARBVs).
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:
NNECO has reviewed the proposed revision in accordance with
10CFR50.92 and has concluded that the revision does not involve a
significant hazards consideration (SHC). The basis for this
conclusion is that the three criteria of 10CFR50.92(c) are not
satisfied. The proposed revision does not involve an SHC because the
revision would not:
1. Involve a significant increase in the probability or
consequence of an accident previously evaluated.
The operability of the SGARBVs provides a method to recover from
a SGTR [steam generator tube rupture] event during which the
operator is required to perform a limited cooldown to establish
adequate subcooling as a necessary step to limit the primary to
secondary break flow into the ruptured steam generator. For other
design events, the SGARBVs provide a safety grade method for cooling
the unit to residual heat removal entry conditions should the
preferred heat sink via the steam bypass system or the steam
generator atmospheric relief valves be unavailable. This proposed
revision to the Technical Specifications will add a new Technical
Specification 3/4.7.1.6 and its associated Bases Section 3/4.7.1.6
which were developed bases on the information contained in the
Westinghouse Improved Standard Technical Specifications, NUREG 1431,
Rev. 1. The proposed specification and bases provide further
assurance that the SGARBVs will be available to function as
described in the accident analysis.
Therefore, the proposed revision does not 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.
This proposed revision to the Technical Specifications to add a
new specification and bases for the SGARBVs does not cause a change
in the operation of any system or component during normal or
accident conditions.
Therefore, the proposed revision does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed new Technical Specification 3/4.7.1.6 and its
associated Bases Section 3/4.7.1.6 were developed based on the
information contained in the Westinghouse Improved Standard
Technical Specifications, NUREG 1431, Rev. 1. The SGARBV's are not
currently in the Technical Specifications of Millstone Unit No. 3
and are being added to ensure accident mitigation functional
capability. The NUREG 1431, Rev. 1 surveillance frequency is 18
months. The NUREG 1431, Rev. 1 surveillance frequency bases reads
``operating experience has shown that these components usually pass
the surveillance when performed at the 18 month frequency''. The
proposed frequency acceptability has been evaluated by reviewing
SGARBV AWO's [automated work order's] for the period from Jan. 1990
to April 1997 to confirm the absence of excessive work orders which
indicate valve functional failures and none were identified.
Additionally, each SGARBV line consists of one SGARBV and an
associated block valve. These proposed changes are consistent with
the design and operation of the SGARBVs. There is no negative affect
on the dose consequences from any design basis event or core damage
frequency.
Therefore, the proposed revision does not involve a significant
reduction in a margin of safety.
In conclusion, based on the information provided, it is
determined that the proposed revision does not involve an SHC.
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: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270
NRC Deputy Director: Phillip F. McKee
Northern States Power Company, Docket Nos. 50-282 and 50-306,
Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue
County, Minnesota
Date of amendment requests: November 27, 1996
Description of amendment requests: The proposed amendment[s] would
incorporate new steam generator tube sleeve designs and installation
and examination techniques into the Prairie Island Technical
Specifications.
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[s] will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The supporting technical evaluation and safety evaluation for
the Combustion Engineering leak tight sleeves demonstrate that the
sleeve configuration will provide steam generator tube structural
and leakage integrity under normal operating and accident
conditions. The sleeve configurations have been designed and
analyzed in accordance with the requirements of the ASME [American
Society of Mechanical Engineers] Code. Mechanical testing has shown
that the sleeve and sleeve joints provide margin above acceptance
[[Page 43371]]
limits. Ultrasonic examination is used to verify the leak tightness
of the above the [sic] tubesheet sleeve welds. Testing has
demonstrated the leak tightness of the hard roll joint as well as
the structural integrity of the hard roll joint. Tube rupture can
not occur at the hard roll joint due to the reinforcing effect of
the tubesheet. Tests have demonstrated that tube collapse will not
occur due to postulated LOCA [loss-of-coolant accident] loadings.
The existing Technical Specification leakage rate requirements
and accident analysis assumptions remain unchanged in the event that
significant leakage did occur from the sleeve joints or that a
sleeve assembly ruptured. Any leakage through the sleeve assembly is
fully bounded by the existing steam generator tube rupture analysis
included in the Prairie Island Plant USAR [updated safety analysis
report]. The proposed sleeving repair does not adversely impact any
other previously evaluated design basis accident.
The sleeve minimum acceptable wall thickness used for developing
the depth based plugging limit for the sleeve is determined using
the guidance of draft Regulatory Guide 1.121 [Bases for
Plugging Degraded PWR [Pressurized-Water Reactor] Steam Generator
Tubes] and the pressure stress equation of Section III of
the ASME Code. Evaluation of the minimum acceptable wall thickness
for normal, upset, and postulated accident condition loading per the
ASME Code finds that the limiting condition is established from
normal operating conditions which then bounds the upset and accident
condition values. Allowance for non-destructive examination and
growth of existing sleeve wall degradation must be made when
determining the sleeve plugging limit. The proposed plugging limit
is 40% through wall degradation. The sleeve assembly will be
examined by state of the art non-destructive examination techniques
on a periodic basis to provide early indication of sleeve
degradation. The corrosion resistance of the Alloy 690 sleeve has
been verified by field experience at Prairie Island. The oldest
Alloy 690 sleeves were installed May 1987. No indication of
corrosion of the sleeve or the parent tube in the weld joint has
been identified by state-of-the-art eddy current examination. These
oldest sleeve welds did not receive post weld heat treatment. In
addition, 5 sleeves were removed for destructive examination in
February, 1996. No corrosion was found in any of these sleeves
including those dating from October 1992. The pulled sleeves had
received post weld heat treatment. Post weld heat treatment can be
optionally applied to the free span sleeve weld joints to reduce the
susceptibility of the weld joint and parent tube to stress corrosion
cracking. Since the sleeve design meets the requirements of the ASME
code and mechanical tests have demonstrated margins above acceptance
criteria, the installation of the Combustion Engineering leak tight
sleeves will not increase the probability or consequences of an
accident previously evaluated.
2. The proposed amendment[s] will not create the possibility of
a new or different kind of accident from any accident previously
analyzed.
Installation of sleeves does not introduce any significant
changes to the plant design basis. The use of a sleeve to span a
degraded region of steam generator tubing restores the structural
and leakage integrity of the tubing to meet the original design
bases. Stress and fatigue analysis of the sleeve assembly shows that
the requirements for ASME Code are met. Mechanical testing has
demonstrated that margin exists above the design criteria. Any
hypothetical accident as a result of any degradation in the sleeved
tube would be bounded by the existing tube rupture accident
analysis.
3. The proposed amendment[s] will not involve a significant
reduction in the margin of safety.
The use of the sleeves to repair degraded steam generator tubing
has been demonstrated to maintain the integrity of the tube bundle
commensurate with the requirements of the ASME Code and draft
Regulatory Guide 1.121 and to maintain the primary to secondary
pressure boundary under normal and postulated accident conditions.
The safety factors used in the verification of the strength of the
sleeve assembly are consistent with the safety factors in the ASME
Boiler and Pressure Vessel Code used in steam generator design. The
operational and faulted condition stresses and cumulative fatigue
usage are bounded by the ASME Code requirements. The sleeve assembly
has been verified by testing to prevent both tube pullout and
significant leakage during normal and postulated accident
conditions. A test program was conducted to ensure the rolled joint
design for the lower joint in the tubesheet sleeve was leak tight
and capable of withstanding the designs loads. The primary coolant
pressure boundary of the sleeve assembly will be periodically
inspected by non-destructive examination to identify sleeve
degradation due to operation. Installation of sleeves will decrease
the number of tubes which must be taken out of service. There is a
small amount of primary coolant flow reduction due to sleeves for
which an equivalent plugging sleeve to plug ratio is assigned and is
used to assess the final equivalent plugging percentage used as an
input to other safety analyses. Because the sleeve maintains the
design basis requirements for the steam generator tubing, it is
concluded that the proposed change does not result in a significant
reduction in margin with respect to plant safety as defined in the
USAR or the Technical Specification Bases.
Based on the evaluation described above, and pursuant to 10 CFR
Part 50, Section 50.91, Northern States Power Company has determined
that operation of the Prairie Island Nuclear Generating Plant in
accordance with the proposed license amendment request does not
involve any significant hazards considerations as defined by NRC
regulations in 10 CFR Part 50, Section 50.92.
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 requests involve 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: Jay Silberg, Esq., Shaw, Pittman, Potts, and
Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: John N. Hannon
Northern States Power Company, Docket Nos. 50-282 and 50-306,
Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue
County, Minnesota
Date of amendment requests: May 15, 1997
Description of amendment requests: The proposed amendments would
change the Technical Specifications (TS) to revise certain limitations
on reactor coolant system leakage and steam generator tube
surveillance. The proposed changes would implement a voltage-based
repair criteria per the requirements of NRC Generic Letter 95-05,
``Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes
Affected by Outside Diameter Stress Corrosion Cracking.'' In addition,
a typographical error in TS Section 4.12.c. is being corrected.
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[s] will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The supporting technical evaluation and safety evaluation for
the voltage based repair criteria demonstrate that steam generator
tube structural and leakage integrity under normal operating and
accident conditions will be maintained. Tube burst criteria are
inherently satisfied during normal operating conditions due to the
proximity of the tube support plate (TSP). Test data referenced in
Generic Letter 95-05 indicates that tube burst cannot occur within
the TSP, even for tubes which have 100% throughwall electric
discharge machining notches, 0.75 inch long, provided that the TSP
is adjacent to the notched area. Since tube-to-TSP proximity
precludes tube burst during normal operating conditions, use of the
criteria must retain tube integrity characteristics which maintain a
margin of safety of 1.43 times the bounding faulted condition, main
steamline break (MSLB) pressure differential. The Regulatory Guide
(RG) 1.121 [Bases for Plugging Degraded PWR [Pressurized-
Water Reactor] Steam Generator Tubes] criterion requiring
maintenance of a safety factor of 1.43 times the MSLB pressure
differential on tube burst
[[Page 43372]]
is satisfied by 7/8'' diameter tubing with bobbin coil indications
with signal amplitudes less than the current 8.7 volts structural
limit, regardless of the indicated depth measurement.
The upper voltage repair limit (VURL) will be
determined prior to each outage using the most recently NRC approved
database to determine the tube structural limit (VSL).
The structural limit is reduced by allowances for nondestructive
examination (NDE) uncertainty (VNDE) and growth
(VGR) to establish VURL. Using the Generic
Letter (GL) 95-05 NDE and growth allowances for an example, the NDE
uncertainty component of 20% and a voltage growth allowance of 30%
per full power year can be utilized to establish a VURL
of 5.2 volts.
Relative to the expected leakage during accident condition
loadings, it has been previously established that a postulated MSLB
outside of containment but upstream of the main steam isolation
valve (MSIV) represents the most limiting radiological conditions to
the plugging criteria. In support of [the] implementation of the
revised plugging limit, analyses will be performed to determine
whether the distribution of cracking indications at the tube support
plate intersections during future cycles are projected to be such
that primary-to secondary leakage would result in postulated off
site and control room doses exceeding the limits established for
application of the voltage-based repair criteria at Prairie Island.
A separate calculation has determined the maximum allowable MSLB
leakage limit in a faulted loop. This limit was calculated using the
technical specification 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 [Standard Review
Plan]. The projected MSLB leak rate calculation
methodology prescribed in Section 2.b of Generic Letter 95-05 will
be used to calculate the end-of-cycle (EOC) leakage. Projected EOC
voltage distribution will be developed using the most recent EOC
eddy current results and considering an appropriate voltage
measurement uncertainty and indication growth allowance. The log-
logistic probability of leakage correlation will be used to
establish the MSLB leak rate used for comparison with the faulted
loop allowable limit. Therefore, as implementation of the voltage-
based repair criteria does not adversely affect steam generator tube
integrity and implementation will be shown to result in acceptable
dose consequences, the proposed amendment[s] [do] not result in any
increase in the probability or consequences of an accident
previously evaluated in the Updated Safety Analysis Report (USAR).
2. The proposed amendment[s] will not create the possibility of
a new or different kind of accident from any accident previously
analyzed.
Implementation of the proposed steam generator tube voltage-
based repair criteria does not introduce any significant changes to
the plant design basis. Use of the voltage-based repair criteria
does not provide a mechanism which could result in an accident
outside of the region of the tube support plate elevations since
tubes with outside diameter stress corrosion cracking (ODSCC) not
occurring inside the thickness of the tube support plates will be
plugged or repaired. Neither a single or multiple tube rupture event
would be expected during all plant conditions in a steam generator
in which the voltage based repair limit has been applied.
Northern States Power will implement a maximum primary-to-
secondary leak rate limit of 150 gpd [gallons per day] per steam
generator to help preclude the potential for excessive leakage
during all plant conditions. The Regulatory Guide 1.121 criterion
for establishing operational leak rate limits that require plant
shutdown are based upon leak-before-break considerations to detect a
free span crack before potential tube rupture during faulted plant
conditions. The 150 gpd limit provides 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.
The operational leakage limit will be reduced to 150 gpd limit
consistent with Generic Letter 95-05. This limit is expected to
provide for plant shutdown prior to reaching critical lengths for
MSLB conditions using the lower 95% leak rate data. 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 and only a small
percentage of the TSPs are deflected greater than the TSP thickness
during a postulated MSLB.
As steam generator tube integrity upon implementation of the
voltage-based repair criteria continues to be maintained through
inservice 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.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed amendment[s] will not involve a significant
reduction in the margin of safety.
The use of the voltage-based repair criteria at Prairie Island
maintains steam generator tube integrity commensurate with the
criteria of the ASME [American Society of Mechanical Engineers] Code
and Regulatory Guide 1.121. Regulatory Guide 1.121 describes a
method acceptable to the Commission for meeting GDCs [General Design
Criteria] 14, 15, 30, 31, and 32 by reducing the probability or the
consequences of steam generator tube rupture. This is accomplished
by determining the limiting conditions of degradation of steam
generator tubing, as established by inservice inspection, for which
tubes with unacceptable cracking should be repaired or removed from
service. Upon implementation of the proposed criteria, even under
the worst case conditions, the occurrence of ODSCC at the tube
support plate elevations is not expected to lead to the steam
generator tube rupture event during normal or faulted plant
conditions. The EOC distribution of crack indications at the tube
support plate elevations will be confirmed to result in acceptable
primary-to-secondary leakage during all plant conditions in order to
assure that radiological consequences meet the requirements of
Generic Letter 95-05.
Previous evaluations have indicated a potential for tube
deformation and collapse during a postulated loss-of-coolant
accident (LOCA) plus safe-shutdown-earthquake (SSE) event. The tube
collapse potential arises from TSP deformation at the support plate
wedges. Evaluation of the Westinghouse umbrella seismic spectra
provided in Westinghouse letter NSP-92-152 for Model 51 steam
generators shows that Prairie Island is bounded by those spectra and
that no tubes will undergo deformation due to the combined effects
of LOCA plus SSE. Therefore, no tubes need to be excluded from
application of the voltage based criteria due to deformation
resulting from combined LOCA plus SSE loadings. Addressing
Regulatory Guide 1.83 [Inservice Inspection of
Pressurized Water Reactor Steam Generator Tubes]
considerations, implementation of the voltage-based repair criteria
is supplemented by enhanced eddy current inspection guidelines to
provide consistency in voltage normalization, by an extensive bobbin
coil inspection which will include 100% of the hot leg TSP
intersections and cold leg intersections down to the lowest cold leg
TSP with known ODSCC, by the determination of the TSPs having ODSCC
using at least 20% random sampling of tubes inspected over their
full length, and by rotating pancake coil inspection (or equivalent)
requirements for the larger indications left in service to
characterize the principal degradation as ODSCC.
As noted previously, implementation of the tube support plate
intersection voltage-based repair criteria will decrease the number
of tubes which must be repaired. The installation of steam generator
tube plugs or sleeves reduces the RCS flow margin. Thus,
implementation of the voltage-based repair 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 with respect to plant safety as defined in the USAR or any
Bases of the plant Technical Specifications.
Based on the evaluation described above, and pursuant to 10 CFR
Part 50, Section 50.91, Northern States Power Company has determined
that operation of the Prairie Island Nuclear Generating Plant in
accordance with the proposed license amendment request does not
involve any significant hazards considerations as defined by NRC
regulations in 10 CFR Part 50, Section 50.92.
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. In addition, the proposed correction to a typographical
error has no effect on the three standards of 10
[[Page 43373]]
CFR 50.92(c). Therefore, the NRC staff proposes to determine that the
amendment requests involve 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: Jay Silberg, Esq., Shaw, Pittman, Potts, and
Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: John N. Hannon
PECO Energy 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 4, 1997
Description of amendment request: The proposed Technical
Specifications (TSs) amendment revises TS Surveillance Requirement
3.8.2.1 to no longer require that automatic emergency diesel generator
(EDG) auto-start and trip bypass features must be functional when the
emergency core cooling system (ECCS) is not required to be operable.
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 to the facility does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The proposed change will eliminate an inconsistency between
Technical Specifications 3.3.5.1, 3.5.2, and 3.8.2 by clarifying
that the EDG auto-start and EDG trip bypass on ECCS initiation
capability is not required during periods in which ECCS is not
required to be OPERABLE. No physical changes to the facility will be
made per this change. The systems, structures, and components
affected by this change are considered to be accident mitigators and
not accident initiators. The affected systems, structures, and
components will continue to operate within the current design
parameters. The ability of the EDGs to auto-start on a loss of
offsite power or degraded voltage will remain unchanged. No new
failure modes or conditions adverse to safety will be created as a
result of this change. Therefore, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. The proposed change to the facility does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed change will eliminate an inconsistency between
Technical Specifications 3.3.5.1, 3.5.2, and 3.8.2 by clarifying
that the EDG auto-start and EDG trip bypass on ECCS initiation
capability is not required during periods in which ECCS is not
required to be OPERABLE. No physical changes to the facility will be
made per this change. The systems, structures and components
affected are considered to be accident mitigators not accident
initiators. The affected systems, structures and components will
continue to operate within the current design parameters. No new
failure modes or conditions adverse to safety will be created as a
result of this change. The plant conditions which do not require any
ECCS to be OPERABLE, (i.e., the plant in MODE 5, the spent fuel
storage pool gates are removed, water level is greater than or equal
to 458 inches above reactor pressure vessel instrument zero, and
there are no OPDRVs [operations with the potential of draining the
reactor vessel] in progress) ensure sufficient coolant inventory to
allow operator action to prevent uncovering the fuel. The ability of
the EDGs to auto-start on a loss of offsite power or degraded
voltage will remain unchanged. Therefore, the proposed change does
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
3. The proposed change to the facility does not involve a
significant reduction in a margin of safety.
The proposed change will eliminate an inconsistency between
Technical Specifications 3.3.5.1, 3.5.2, and 3.8.2 by clarifying
that the EDG auto-start and EDG trip bypass on ECCS initiation
capability is not required during periods in which ECCS is not
required to be OPERABLE. The ECCS and EDGs capability to perform the
required safety functions as described/required in the bases of the
current plant Technical Specifications will be maintained.
Therefore, the proposed change to the facility does not result in a
significant reduction in any 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,
PA 17105
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
PA 19101
NRC Project Director: John F. Stolz
PECO Energy Company, Public Service Electric and Gas Company,
Delmarva Power and Light Company, and Atlantic City Electric
Company, Docket No. 50-278, Peach Bottom Atomic Power Station, Unit
No. 3, YorkCounty, Pennsylvania
Date of application for amendment: June 30, 1997
Description of amendment request: The proposed amendment would
revise the Technical Specification (TS) 2.1.1.2 safety limit minimum
critical power ratios (SLMCPRs) to be consistent with the use of GE 13
fuel in the Unit 3 core for operating cycle 12.
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 TS changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The derivation of the cycle-specific SLMCPRs for incorporation
into the TS, and its use to determine cycle-specific thermal limits,
have been performed using the methodology discussed in ``General
Electric Standard Application for Reactor Fuel,'' NEDE-24011-P-A-13,
and U.S. Supplement, NEDE-24011-P-A-13-US, August, 1996, and the
``Proposed Amendment 25 to GE Licensing Topical Report NEDE-24011-P-
A (GESTAR II) on Cycle Specific Safety Limit MCPR.'' Amendment 25
was submitted by GENE to the U.S. Nuclear Regulatory Commission
(USNRC) on December 13, 1996. This change in SLMCPRs cannot increase
the probability or severity of an accident.
The basis of the SLMCPR calculation is to ensure that greater
than 99.9% of all fuel rods in the core avoid transition boiling if
the limit is not violated. The new SLMCPRs preserve the existing
margin to transition boiling and fuel damage in the event of a
postulated accident. The fuel licensing acceptance criteria for the
SLMCPR calculation apply to PBAPS, Unit 3, Cycle 12 in the same
manner as they have applied previously. The probability of fuel
damage is not increased. 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.
The SLMCPR is a TS numerical value, designed to ensure that
transition boiling does not occur in 99.9% of all fuel rods in the
core during the limiting postulated accident. It cannot create the
possibility of any new type of accident. The new SLMCPRs are
calculated using methodology discussed in ``General Electric
Standard Application for Reactor Fuel,'' NEDE-24011-P-A-13, and U.S.
Supplement, NEDE-24011-P-A-13-US, August, 1996, and the ``Proposed
[[Page 43374]]
Amendment 25 to GE Licensing Topical Report NEDE-24011-P-A (GESTAR
II) on Cycle Specific Safety Limit MCPR.'' Amendment 25 was
submitted by GENE to the U.S. Nuclear Regulatory Commission (USNRC)
on December 13, 1996.
Therefore, the proposed TS changes do not create the possibility
of a new or different kind of accident, from any accident previously
evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The margin of safety as defined in the TS Bases will remain the
same. The new SLMCPRs are calculated using methodology discussed in
``General Electric Standard Application for Reactor Fuel,'' NEDE-
24011-P-A-13, and U.S. Supplement, NEDE-24011-P-A-13-US, August,
1996, and the ``Proposed Amendment 25 to GE Licensing Topical Report
NEDE-24011-P-A (GESTAR II) on Cycle Specific Safety Limit MCPR.''
Amendment 25 was submitted by GENE to the U.S. Nuclear Regulatory
Commission (USNRC) on December 13, 1996. The fuel licensing
acceptance criteria for the calculation of the SLMCPR apply to PBAPS
[Peach Bottom Atomic Power Station], Unit 3 Cycle 12 in the same
manner as they have applied previously. The SLMCPRs ensure that
greater than 99.9% of all fuel rods in the core will avoid
transition boiling if the limit is not violated, thereby preserving
the fuel cladding integrity. 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: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
PA 17105
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
PA 19101
NRC Project Director: John F. Stolz
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: April 14, 1997
Description of amendment request: The proposed amendment revises
Appendix A, Section 6 of the Technical Specifications. The changes will
enable Safety Review Committee (SRC) to review plant staff performance
by deleting the plant staff performance requirement from Section
6.5.2.9.b and incorporating a plant staff review requirement in Section
6.5.2.8. The amendment also replaces the position title of Vice
President (VP) Regulatory Affairs and Special Projects (RASP) with
Director of Regulatory Affairs and Special Projects.
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 analyzed?
Response:
This amendment application does not involve a significant
increase in the probability or consequences of an accident
previously analyzed. The proposed changes allow the SRC to perform a
review, rather than an audit, of plant staff performance. This
change does not diminish the SRC's effectiveness. A review of the
1995 QA [quality assurance] audit of plant staff performance shows
that no findings related to plant staff performance were issued.
This indicates that the other review mechanisms currently in place
are sufficient to ensure that plant staff performance is monitored.
The position title change of VP-RASP to Director-RASP is an
administrative change as all previously performed functions are
being maintained. Therefore, the proposed changes do not affect the
probability or consequences of any previously analyzed accident.
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any accident previously
evaluated?
Response:
This amendment application does not create the possibility of a
new or different kind of accident from any accident previously
evaluated. The proposed changes affect an SRC audit requirement and
a management position title. These changes do not affect plant
equipment or the way the plant operates. Therefore, they cannot
create a new or different kind of accident.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
Response:
This amendment application does not involve a significant
reduction in a margin of safety. The requested Technical
Specification revisions require the SRC to review rather than audit
facility staff performance and will not diminish the effectiveness
of the SRC. A review of the 1995 audit confirms that performance of
the annual audit is redundant as no findings or recommendations
concerning plant staff performance were made. The QA/ORG quarterly
trend reports and SRC review of facility staff performance are
adequate to ensure that plant staff performance is properly
monitored.
The position title change (VP-RASP to Director-RASP) is an
administrative change as all previously performed functions are
being maintained. 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 proposed 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: Alexander W. Dromerick, Acting Project
Director
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of amendment request: March 31, 1997, as supplemented by
letter dated July 16, 1997. The July 16, 1997, supplement supersedes
the March 31, 1997 application.
Description of amendment request: The proposed amendment would
provide changes to Technical Specification (TS) 2.1.2, ``THERMAL POWER,
High Pressure and High Flow,'' ACTION a.1.c for TS 3.4.1.1,
``Recirculation Loops,'' and the Bases for TS 2.1, ``Safety Limits.''
These changes are being made to implement an appropriately conservative
Safety Limit Minimum Critical Power Ratio, to include Cycle 8 specific
analyses, for all Hope Creek core and fuel designs.
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 derivation of the revised SLMCPRs for Hope Creek for
incorporation into the Technical Specifications, and its use to
determine cyclespecific thermal limits, have been performed using
NRC approved methods. Additionally, interim implementing procedures
which incorporate cyclespecific parameters have been used which
result in a more restrictive value for SLMCPR. These calculations do
not change the method of operating the plant and have no effect on
the probability of an accident initiating event or transient.
There are no significant increases in the consequences of an
accident previously evaluated. The basis of the MCPR Safety Limit is
to ensure that no mechanistic fuel damage is calculated to occur if
the limit is not violated. The new SLMCPRs preserve the
[[Page 43375]]
existing margin to transition boiling and the probability of fuel
damage is not increased. Therefore, the proposed change does 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 accident previously
evaluated.
The proposed changes contained in this submittal result from an
analysis of the Cycle 7 and Cycle 8 core reloads using the same fuel
types as previous cycles. These changes do not involve any new
method for operating the facility and do not involve any facility
modifications. No new initiating events or transients result from
these changes. Therefore, the proposed Technical Specification
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 margin of safety as defined in the Technical Specification
bases will remain the same. The new SLMCPRs are calculated using NRC
approved methods which are in accordance with the current fuel
design and licensing criteria. Additionally, interim implementing
procedures, which incorporate cyclespecific parameters, have been
used. The MCPR Safety Limit remains high enough to ensure that
greater than 99.9% of all fuel rods in the core will avoid
transition boiling if the limit is not violated, thereby preserving
the fuel cladding integrity. Therefore, the proposed Technical
Specification 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: Pennsville Public Library, 190
S. Broadway, Pennsville, NJ 08070
Attorney for licensee: J. J. Keenan, Esquire, Nuclear Business Unit
- N21, P.O. Box 236, Hancocks Bridge, NJ 08038
NRC Project Director: John F. Stolz
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of amendment request: April 1, 1997, as supplemented by letter
dated May 30, 1997
Description of amendment request: The proposed amendment would
provide changes to Technical Specifications (TSs) 4.6.1.1, ``Primary
Containment Integrity,'' 3/4.6.1.2, ``Primary Containment Leakage,'' 3/
4.6.1.3, ``Primary Containment Air Locks,'' 4.6.1.5.1, ``Primary
Containment Structural Integrity,'' and 4.6.1.8.2, ``Drywell and
Suppression Chamber Purge System.'' The amendment would also change the
Bases for 3/4.6.1.2, ``Primary Containment Leakage,'' 3/4.6.1.3,
``Primary Containment Air Locks,'' 3.4.6.1.5, ``Primary Containment
Structural Integrity,'' Section 6, ``Administrative Controls,'' and
License Condition 2.D of Facility Operating License NPF-57. A new TS,
6.8.4.e, ``Primary Containment Leakage Rate Testing Program,'' would be
added. These changes modify the TSs and the Facility Operating License
to adopt the performance based containment leak rate testing
requirements (Option B) of 10 CFR Part 50, Appendix J.
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.
Containment leak rate testing is not an initiator of any
accident. The proposed changes do not make any physical changes to
the containment and do not affect reactor operations or the accident
analyses. Therefore, the proposed changes do not involve a
significant increase in the probability of any previously evaluated
accident.
Since the allowable leakage rate is not being changed and since
the analysis documented in NUREG-1493, ``Performance-Based
Containment Leak-Test Program'' concludes that the impact on public
health and safety due to extended intervals is negligible, the
proposed changes will not involve a significant increase in the
consequences of any previously evaluated accident.
Therefore, adoption of a performance-based leakage testing
requirements will provide an equivalent level of safety and does 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.
No physical changes are being made to the plant, nor are there
any changes being made to the operation of the plant as a result of
the proposed changes. In addition, no new failure modes of plant
equipment previously evaluated are being introduced.
Therefore, the proposed amendment will 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 are based on NRC-accepted provisions and
maintain adequate levels of reliability of containment integrity.
The performance-based approach to leakage rate testing recognizes
that historically good results of containment testing provide
appropriate assurance of future containment integrity. This supports
the conclusion that the impact on the health and safety of the
public as a result of extended test intervals is negligible. Since
the analysis documented in NUREG-1493 confirms that the performance
based schedule continues to maintain a minimal impact on public
risk, it can be concluded that the margin of safety is not
significantly affected by the proposed changes.
Therefore, the proposed amendment 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: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit - N21, P. O. Box 236, Hancocks Bridge, New Jersey 08038
NRC Project Director: John F. Stolz
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of amendment request: July 3, 1997
Description of amendment request: The proposed amendment would
change Technical Specification Table 3.6.3-1, ``Primary Containment
Isolation Valves'' to add valves to the list, therein.
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 accidents previously evaluated in the UFSAR [Updated Final
Safety Analysis Report] that could be possibly affected by this
proposal are those involving loss of coolant scenarios such as a
piping or instrument line break. The proposed relief valves,
associated piping and the affected portions of containment
penetration piping are not initiators of those accidents evaluated
in the UFSAR. The proposed relief valves limit the post-accident
maximum expected pressures of the affected piping segments within
ASME [American Society of Mechanical Engineers] code allowables and
system design pressures. The modification does not cause any system
or component to be operated outside of their design rating
[[Page 43376]]
allowed by applicable codes. The proposed relief valves will be
safety-related and Seismic Category I components (except for the
relief valve discharge piping, which will be non-safety related and
seismically analyzed, and will meet the design, material and
construction standards applicable to the affected piping
segments[)].
The proposed modifications do not jeopardize the capability of
the containment isolation valves in the affected penetrations to
close on the receipt of a containment isolation signal or to
mitigate the consequences of design basis accidents evaluated in the
UFSAR. Although the modifications will result in system pressures to
be above their currently established design values, the new peak
operating pressures of the affected piping segments will be limited
to within the requirements of the ASME code. The modification will
not alter any assumptions previously made or change, degrade, or
prevent actions described in or assumed in evaluating the
radiological consequences of the postulated design basis accidents.
Containment structure temperature and pressure limits will not be
exceeded with this modification and the offsite dose consequences
will not be affected.
Therefore these changes will not significantly increase the
probability of an accident previously evaluated, nor involve a
significant increase in the 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.
Accidents or malfunctions of equipment important to safety
previously evaluated in the UFSAR relating to the proposed
modification involve the single active failure of a containment
isolation valve to close upon receipt of a containment isolation
signal or its failure to limit the containment bypass leakage
following its closure. The proposed modification: 1) does not impact
the automatic closure times of the containment isolation valves; 2)
does not impact their capability to maintain leak tightness during a
postulated design basis accident; and 3) does not adversely impact
the manner in which any system is operated. The proposed
modification does not compromise the UFSAR accident analysis
assumptions and/or limits. The licensing basis safety analysis
limits for all systems important to safety continue to be met.
Furthermore, there is no change in plant testing proposed in this
change request which could initiate an event. Therefore, these
changes will 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 proposed modifications and Technical Specification changes
do not change the design limits, acceptance criteria or accident
analysis assumptions pertaining to the containment isolation valves,
their associated piping or any other safety-related systems,
structures or components. The proposed modification does not impact
the automatic closure times of the containment isolation valves, nor
does it impact their capability to maintain leak tightness during a
postulated design basis accident. For the systems affected by these
penetration modifications, there is no change in system function or
structural integrity introduced with these proposed changes.
Therefore, the changes contained in this request do not result in a
significant reduction in a margin of safety for the containment
isolation capability of Hope Creek.
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: Pennsville Public Library, 190
S. Broadway, Pennsville, NJ 08070
Attorney for licensee: J. J. Keenan, Esquire, Nuclear Business Unit
- N21, P.O. Box 236, Hancocks Bridge, NJ 08038
NRC Project Director: John F. Stolz
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of amendment request: July 7, 1997
Description of amendment request: The proposed amendment would
change Technical Specification (TS) 3/4.8.4.2, ``Motor Operated Valves
- Thermal Overload Protection (BYPASSED),'' to relocate the list of
applicable valves (TS Table 3.8.4.2-1) to the Hope Creek (HC)
Generating Station Updated Final Safety Analysis Report (UFSAR).
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 proposed TS revisions involve: 1) no hardware changes; 2) no
changes to the operation of any systems or components in normal or
accident operating conditions; and 3) no changes to existing
structures, systems or components. The relocation of Technical
Specification Table 3.8.4.2-1 to the UFSAR and existing surveillance
procedures will continue to ensure that safety-related motor-
operated valves (MOVs) are capable of performing their intended
safety functions. Therefore these changes will not significantly
increase the probability of an accident previously evaluated. To the
extent practicable, these proposed changes were developed consistent
with the changes approved by the NRC when developing NUREG-1433,
``Standard Technical Specifications, General Electric Plants, BWR/
4'', with the intent of having this relocated information controlled
in other plant documents subject to 10CFR50.59 provisions. Since the
plant systems associated with these proposed changes will still be
capable of: 1) meeting all applicable design basis requirements; and
2) retain the capability to mitigate the consequences of accidents
described in the HC UFSAR, the proposed changes were determined to
be justified. Therefore, these changes will not involve a
significant increase in the 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.
Relocation of Technical Specification Table 3.8.4.2-1 to the
UFSAR will not adversely impact the operation of any safety related
component or equipment. Since the proposed changes involve: 1) no
hardware changes; 2) no changes to the operation of any systems or
components; and 3) no changes to existing structures, systems or
components, there can be no impact on the occurrence of any
accident. To the extent practicable, these proposed changes were
developed consistent with the changes approved by the NRC when
developing NUREG-1433, ``Standard Technical Specifications, General
Electric Plants, BWR/4'', with the intent of having this relocated
information controlled in other plant documents subject to
10CFR50.59 provisions. Furthermore, there is no change in plant
testing proposed in this change request which could initiate an
event. Therefore, these changes will 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.
Relocation of Technical Specification Table 3.8.4.2-1 to the
UFSAR is consistent, to the extent practicable, with the changes
approved by the NRC when developing NUREG-1433, ``Standard Technical
Specifications, General Electric Plants, BWR/4''. The MOV thermal
overload protection table will reside in the UFSAR and will ensure
that the associated MOVs will be capable of performing their
intended safety functions. Any changes to this UFSAR table will be
subject to the provisions of 10CFR50.59 and a separate safety
evaluation would be developed to support any proposed changes that
would subsequently be made. Therefore, the changes contained in this
request do not result in 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: Pennsville Public Library, 190
S. Broadway, Pennsville, NJ 08070
Attorney for licensee: J. J. Keenan, Esquire, Nuclear Business Unit
- N21,
[[Page 43377]]
P.O. Box 236, Hancocks Bridge, NJ 08038
NRC Project Director: John F. Stolz
Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County,
Alabama
Date of amendment request: June 2, 1997 (TS 387)
Description of amendment request: The proposed amendment allows
continued plant operation with a single reactor recirculation loop in
service. The Nuclear Regulatory Commission has previously determined
single loop operation is generically acceptable as set forth in Generic
Letter 86-09, ``Technical Resolution of Generic Issue B-59-(N-1) Loop
Operation in BWRs [boiling water reactors] and PWRs [pressurized-water
reactors].'' Single loop operation is also recognized as a standard
mode of operation in the BWR/4 Improved Standard TS.
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 or consequences of an accident
previously evaluated.
An analysis of the limiting operational transients has been
performed by GE [General Electric] for BFN as documented in NEDO-
24236 to demonstrate adequate margin to the Safety Limit Minimum
Critical Power Ratio (SLMCPR). In addition, SLO [single loop
operation] has been specified as a operating option for the
transient and accident evaluations performed as part of the cycle-
specific core reload analyses for Units 2 and 3 which ensure that
operating limit Minimum Critical Power Ratios (OLMCPRs) for the
current fuel types are established that maintain required margin to
the fuel cladding safety limit. A cycle-specific analysis with SLO
will be performed for Unit 1 prior to restart and experience
indicates similar results are expected as those for Units 2 and 3.
A review of the values used in the statistical analysis used in
the basis of the fuel cladding safety limit determined that, due to
increased uncertainties in total core flow readings and Traversing
In-Core Probe (TIP) readings during SLO, an increase in the SLMCPR
of .02 is bounding when in SLO. Therefore, while operating in
single-loop mode, an additional .02 is added to the OLMCPR which
maintains the same margin to the fuel cladding safety limit as that
established for two-loop operation. This is a conservative approach
because the two-loop transients have been shown to be more severe
than the equivalent single-loop events and, therefore, the OLMCPRs
established for two-loop operation would always be bounding. Thus,
the margin of safety for fuel clad integrity is assured and the
probability or consequences associated with reactor transients is
not increased for SLO.
SLO results in backflow through the jet pumps in the inactive
recirculation loop which perturbs the relationship between the core
flow and recirculation drive flow on which the flow biased Average
Power Range Monitor (APRM) and Rod Block Monitor (RBM) setpoint
equations are based. To compensate, the proposed TS [Technical
Specification] changes modify the setpoint equations to correct for
one-loop operation. With this adjustment, the setpoint equations
preserve the original relationship between the setpoints and the
effective recirculation drive flow such that the consequences of a
RWE [rod withdrawal event] in SLO are bounded by the cycle-specific
RWE analyses. Therefore, these changes do not increase the
probability or consequences of the RWE transient previously
evaluated.
Average Planar Linear Heat Generation Rate (APLHGR) limits are
established to ensure the acceptance criteria for fuel and Emergency
Core Cooling Systems established in 10 CFR 50.46 are met. A SLO Loss
of Coolant Accident (LOCA) analysis was performed using the SAFER/
GESTR computer code as documented in NEDC-32484P, Revision 1,
``Browns Ferry Nuclear Plant, Units 1, 2, and 3, SAFER/GESTR-LOCA,
Loss-of-Coolant Accident Analysis.''
The LOCA [loss of cooling accident] results for SLO using SAFER/
GESTR showed that, with the application of an APLHGR multiplier as
proposed in the TS change, the LOCA peak clad temperature for SLO
will always be lower than that for limiting design basis pipe break
for two-loop operation. An APLHGR multiplier of 0.9 is applicable
for all current fuel types being used. This multiplier is documented
in each cycle-specific reload analysis and included in the COLR
[core operating limits report]. NEDC-32484P Revision 1 also
concludes that the design basis accident (large breaks) are more
affected than small break sequences and, therefore, the large break
results are bounding for SLO.
The Recirculation Pump Seizure event in SLO was evaluated in
NEDO-24236 and shown to be a non-limiting event. This conclusion is
also supported by GE analyses on other BWRs.
In summary, based on the above discussion, the proposed changes
for SLO do not increase the probability or consequences of an
accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Although the proposed change allows extended operation in a
configuration that was previously allowed for a limited period,
analysis has shown (as described in item A above), that operation
with one recirculation pump out-of-service is within existing
analyses based on the proposed TS requirements. Therefore, the
proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed change to operate in single-loop recirculation mode
has been analyzed in accordance with established transient and
accident methodologies, and margins of safety for the design basis
accidents and transients analyzed in Chapter 14 of the BFN UFSAR
[updated final safety analysis report] have not been significantly
reduced. The basis for this conclusion is outlined in item A above.
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 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: Athens Public Library, 405 E.
South Street, Athens, Alabama 35611
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271,
Vermont Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: June 9, 1997
Description of amendment request: The amendment proposes to update
the Technical Specifications, Section 6.0, to add a reference to NRC-
approved methodologies which will be used to validate or generate the
operating limits in the Vermont Yankee Core Operating Limits 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 change will not involve any significant increase
in the probability or consequences of an accident. The change
updates the Technical Specifications to include [an] NRC approved
method reference to allow calculation of thermal hydraulic stability
limits. It does not affect plant operation and will not weaken or
degrade the facility.
2. The proposed change will not create the possibility of a new
or different kind of accident since the change is administrative. No
physical alterations of the plant, setpoint changes, or operating
conditions are proposed.
3. The proposed change will not involve a significant reduction
in a margin of safety. The change involves an update to the
Administrative Controls in Section 6.0 of the Technical
Specifications by adding a reference to NRC approved methods. This
administrative change does not alter plant safety margins.
[[Page 43378]]
The NRC staff has reviewed the licensee's analysis and, based
onthis 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: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301
Attorney for licensee: R. K. Gad, III, Ropes and Gray, One
International Place, Boston, MA 02110-2624
NRC Project Director: Ronald B. Eaton, Acting
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.
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 amendment request: June 9, 1997
Description of amendment request: The proposed amendments authorize
a revision to the realistic dose values for the process gas system
rupture in Section 15.0 of the Byron/Braidwood (B/B) Updated Final
Safety Analysis Report (UFSAR). During preparation of a UFSAR change
package, ComEd discovered that the Final Safety Analysis Report (FSAR)
had not been updated to correct an error from the previous revision of
the dose calculation. Since the correct dose value is greater than that
previously reported, the consequences of the accident had increased,
and an unreviewed safety question resulted.
Date of publication of individual notice in Federal Register: July
10, 1997 (62 FR 37079).
Expiration date of individual notice: August 11, 1997 (as corrected
(62 FR 39282)).
Local Public Document Room location: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of amendment request: June 27, 1997, as supplemented by letter
dated July 2, 1997 The supplemental letter provided clarifying
information and did not change the initial proposed no significant
hazards consideration determination.
Brief description of amendment request: These amendments clarify,
in the technical specifications (TSs) for each unit, the methodology
used to satisfy surveillance requirements for the laboratory analysis
of activated carbon (charcoal) samples from the standby gas treatment
system (SGTS) and the control room emergency outside air supply system
(CREOASS). The specific changes are made to Sections 4.6.5.3.b.2 and
4.6.5.3.c for the SGTS and to Sections 4.7.b.2 and 4.7.2.c for the
CREOASS, to include a reference to American Society for Testing
Materials (ASTM), ``Radioiodine Testing of Nuclear-Grade Gas Phase
Adsorbents,'' ASTM D3803-79. Date of publication of individual notice
in Federal Register: July 8, 1997 (62 FR 36580)
Expiration date of individual notice: August 7, 1997
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701
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, and at the local public document rooms for
the particular facilities involved.
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: April 14, 1997
Brief description of amendments: The amendments revise Technical
Specification (TS) 3/4.3.8, ``Feedwater/Main Turbine Trip System
Actuation Instrumentation'' by changing the minimum channels required
from three to four. This change reflects a modification that is being
installed to add an auxiliary contact to the trip system logic. In
addition, the amendments revise the TS action statement for inoperable
channels to be consistent with the Improved Standard Technical
Specifications and to account for the additional channel.
Date of issuance: July 29, 1997
Effective date: Immediately, to be implemented within 60 days.
Amendment Nos.: 119 and 104
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 18, 1997 (62 FR
33120). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 29, 1997. No significant
hazards consideration comments received: No.
[[Page 43379]]
Local Public Document Room location: Jacobs Memorial Library,
Illinois Valley Community College, Oglesby, Illinois 61348
Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van
Buren County, Michigan
Date of application for amendment: March 27, 1997, as supplemented
July 7, 1997
Brief description of amendment: The amendment revises the Palisades
Plant license and technical specifications to reflect the licensee's
name change from ``Consumers Power Company'' to ``Consumers Energy
Company.''
Date of issuance: July 21, 1997
Effective date: July 21, 1997
Amendment No.: 176
Facility Operating License No. DPR-20: Amendment revised the
license and the technical specifications.
Date of initial notice in Federal Register: April 23, 1997 (62 FR
19828) The July 7, 1997, letter provided supplementary information
within the scope of the original application and did not change the NRC
staff's initial proposed no significant hazards considerations
determination. The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 21, 1997. No significant
hazards consideration comments received: No.
Local Public Document Room location: Van Wylen Library, Hope
College, Holland, Michigan 49423
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: May 27, 1997
Brief description of amendments: The amendments delete Section
4.7.13.3.a.2 of each unit's Technical Specifications, regarding the
minimum volume and boron concentration of borated water available to
the Standby Makeup Pump of the Standby Shutdown System.
Date of issuance: July 21, 1997
Effective date: As of the date of issuance to be implemented within
30 days
Amendment Nos.: 160 and 152
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 18, 1997 (62 FR
33121) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 21, 1997. 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 Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: February 17, 1997, as revised
May 1, 1997.
Brief description of amendment: Changes to Technical Specification
(TS) to implement 10 CFR 50, Appendix J Option B relating to
containment leakage tests.
Date of issuance: July 24, 1997
Effective date: July 24, 1997
Amendment No.: 156
Facility Operating License No. DPR-72: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 28, 1997 (62
FR 9214), as superseded June 4, 1997 (62 FR 30632) The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated July 24, 1997. No significant hazards consideration comments
received: No.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 32629
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: April 28, 1997
Brief description of amendment: Technical Specification (TS) 3.7.6
requires that flood protection be provided for the service water pump
cubicles and components when the water level exceeds a specific value.
The amendment (1) adds the closing of the service water pump cubicle
sump drain valves to the TS, (2) revises the wording of the action
statement to be consistent with the limiting condition for operation,
and (3) revises the associated Bases section.
Date of issuance: July 28, 1997
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 144
Facility Operating License No. NPF-49: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 4, 1997 (62 FR
30636) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 28, 1997. No significant
hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut 06360, and the Waterford Library, ATTN: Vince
Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385
Northern States Power Company, Docket No. 50-263, Monticello
Nuclear Generating Plant, Wright County, Minnesota
Date of application for amendment: January 23, 1997, as
supplemented January 28, March 4, June 19, July 2, July 16 (2 letters),
July 21, and July 25, 1997
Brief description of amendment: The amendment documents the staff's
review and approval of the apparent unreviewed safety questions (USQs)
associated with (1) the updated analysis of the design-basis accident
(DBA) containment temperature and pressure response, and (2) the
reliance on containment pressure to compensate for the potential
deficiency in net positive suction head (NPSH) for the emergency core
cooling system (ECCS) pumps during a DBA with the worst case scenario
assumptions. The amendment also authorizes the licensee to change the
Technical Specification bases and the Updated Safety Analysis Report,
to reflect the reliance of containment pressure to compensate for the
potential deficiency in NPSH for the ECCS pumps following a DBA.
Date of issuance: July 25, 1997
Effective date: July 25, 1997. Implementation shall be as specified
in Appendix C to the license.
Amendment No.: 98
Facility Operating License No. DPR-22: Amendment revised the
license and the licensee's updated safety analysis report.
Date of initial notice in Federal Register: February 12, 1997 (62
FR 6576) The June 19, 1997, submittal, expanded the scope of the
initial submittal dated January 23, 1997, and therefore, another notice
was issued in Federal Register on June 24, 1997 (62 FR 34086). The July
2, July 16 (2 letters), July 21, and July 25, 1997, submittals provided
additional clarifying information within the scope of the application
and did not change the NRC staff's proposed no significant hazards
considerations determination that was based on the June 19, 1997,
submittal. Therefore, renoticing was not warranted. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated July 25, 1997. No significant hazards consideration comments
received: No.
Local Public Document Room location: Minneapolis Public Library,
[[Page 43380]]
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of application for amendments: June 27, 1997, as supplemented
by letter dated July 2, 1997 The supplemental letter provided
clarifying information and did not change the initial proposed no
significant hazards consideration determination.
Brief description of amendments: These amendments clarify, in the
technical specifications (TSs) for each unit, the methodology used to
satisfy surveillance requirements for the laboratory analysis of
activated carbon (charcoal) samples from the standby gas treatment
system (SGTS) and the control room emergency outside air supply system
(CREOASS). The specific changes are made to Sections 4.6.5.3.b.2 and
4.6.5.3.c for the SGTS and to Sections 4.7.b.2 and 4.7.2.c for the
CREOASS, to include a reference to American Society for Testing
Materials (ASTM), ``Radioiodine Testing of Nuclear-Grade Gas Phase
Adsorbents,'' ASTM D3803-79.
Date of issuance: July 30, 1997
Effective date: Both units, as of date of issuance, to be
implemented within 30 days.
Amendment Nos.: 167 and 141
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the Technical Specifications. Public comments requested as to
proposed no significant hazards consideration: Yes (62 FR 36580). That
notice provided an opportunity to submit comments on the Commission's
proposed no significant hazards consideration determination by July 22,
1997. No comments have been received. The notice also provided an
opportunity to request a hearing by August 7, 1997, 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. On July 9, 1997, the NRC staff issued a Notice of
Enforcement Discretion in order to delay enforcement of the current,
subject, TS requirements until the NRC could take formal action on the
July 2, 1997, application. The Commission's related evaluation of the
amendments, finding of exigent circumstances, consultation with the
State of Pennsylvania, and final no significant hazards consideration
determination are contained in a Safety Evaluation dated July 30, 1997.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of application for amendment: February 11, 1997.
Brief description of amendment: This amendment changes the Hope
Creek Technical Specification (TS) Sections 3/4.8.1, ``A.C. Sources,''
6.8, ``Procedures and Programs,'' and the Bases for Section 3/4.8,
``Electrical Power Systems,'' to include: 1) the relocation of existing
surveillance requirements related to diesel fuel oil chemistry; 2) the
introduction of a new program under TS 6.8.4.e, ``Diesel Fuel Oil
Testing Program; 3) revisions to the TS Bases for Section 3/
4.8 to incorporate information associated with the TS changes; and 4)
editorial changes to implement required corrections.
Date of issuance: July 24, 1997
Effective date: As of date of issuance, to be implemented within 60
days.
Amendment No.: 100
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 26, 1997 (62 FR
14469) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 24, 1997. No significant
hazards consideration comments received: No.
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of application for amendment: March 3, 1997, as supplemented
by letter dated May 5, 1997
Brief description of amendment: This amendment changes Hope Creek
TSs as follows: (1) TS 3/4.3.1, ``Reactor Protection System
Instrumentation,'' TS 3/4.3.2, ``Isolation Actuation Instrumentation,''
and TS 3/4.3.3, ``Emergency Core Cooling System Actuation
Instrumentation,'' to include additional information concerning
response time testing; (2) TS 4.0.5 to reference inservice inspection
and test requirements; (3) TS 3/4.6.1, ``Primary Containment,'' and
associated Bases to reflect a design modification; (4) TS 3/4.7.7,
``Main Turbine Bypass System,'' to specify a new operability
requirement; and (5) the Bases for TS 3/4.8, ``Electrical Power
Systems.''
Date of issuance: July 24, 1997
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 101
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 18, 1997 (62 FR
33131) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 24, 1997. No significant
hazards consideration comments received: No.
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
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 application for amendments: February 11, 1997, as
supplemented on May 1, June 12, and July 23, 1997
Brief description of amendments: The amendments add a new Technical
Specification, 3/4.7.10, ``Chilled Water System - Auxiliary Building
Subsystem,'' and an associated Bases section to address the support
function this system provides to other necessary safety systems.
Date of issuance: July 29, 1997
Effective date: Unit 1 to be implemented prior to entering Mode 6
from the current unit outage; Unit 2 as of its date of issuance, to be
implemented within 10 days of issuance.
Amendment Nos.: 199 and 182
Facility Operating License Nos. DPR-70 and DPR-75.: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 12, 1997 (62 FR
11497) The licensee's supplemental letters provided additional
information that did not affect the staff's proposed no significant
hazards consideration determination. The Commission's related
evaluation of the amendments is contained in a Safety Evaluation dated
July 29, 1997. No significant hazards consideration comments received:
No.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear
Plant, Unit 1, Rhea County, Tennessee
Date of application for amendment: October 23, 1996, as
supplemented
[[Page 43381]]
December 11, 1996, January 31, February 10 and 24, March 11, April 4
and 11, May 28, June 26, and July 15, 1997.
Brief description of amendment: The amendment changes the Watts Bar
Nuclear Plant, Unit 1, Technical Specifications (TS) to increase the
spent fuel storage capacity from 484 fuel assemblies to 1610 fuel
assemblies and to increase the initial enrichment of the fuel to be
stored in the spent fuel storage racks from 3.5 weight percent (wt%) to
5.0 wt%. This modification also changes the center-to-center spacing of
stored fuel assemblies and reflects the use of burnup credit rack
modules to be installed peripherally along the pool walls.
The amendment, as proposed by the licensee, would also involve the
installation of spent fuel racks in the spent fuel cask pit for 225
storage spaces thus increasing the total WBN spent fuel storage
capacity to 1835 spent fuel assemblies. The licensee proposed to
provide an impact shield that would be placed over the fuel in the cask
pit when heavy loads are moved near or across the cask pit area. The
staff is continuing its review of this aspect of the licensee's
proposal. Accordingly, this amendment authorizes the reracking and
usage of the main spent fuel pool, as proposed for a total of 1610
spent fuel spaces. However, it does not authorize the installation of
storage racks or storage of spent fuel in the spent fuel cask pit. The
staff's review of that aspect of the licensee's application will be
addressed by further correspondence.
Date of issuance: July 28, 1997
Effective date: July 28, 1997
Amendment No.: 6
Facility Operating License No. NPF-90: Amendment revises the TS.
Date of initial notice in Federal Register: April 2, 1997 (62 FR
15733) The April 4, and 11, May 28, June 26 and July 15, 1997 letters
provided clarifying informaion that did not change the initial proposed
no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in an environmental assessment dated April 7, 1997, and a Safety
Evaluation dated July 28, 1997. No significant hazards consideration
comments received: None
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, TN 37402
Virginia Electric and Power Company, et al., Docket Nos. 50-338 and
50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa
County, Virginia
Date of application for amendments: November 9, 1987, as
supplemented March 31, 1988, June 8, 1992, and February 4, 1997
Brief description of amendments: These amendments reformat the
operability and surveillance requirements for the intermediate range
channels.
Date of issuance: July 30, 1997
Effective date: July 30, 1997
Amendment Nos.: 206 and 187
Facility Operating License Nos. NPF-4 and NPF-7: Amendments revised
the Technical Specifications.
Date of initial notice in Federal Register: June 18, 1997 (62 FR
33136) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 30, 1997. No significant
hazards consideration comments received: No.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: February 17, 1997
Brief description of amendment: The amendment revises the technical
specifications to move Table 3.6-1, ``Containment Isolation Valves'' to
Wolf Creek Generating Station procedures. In addition, the technical
specifications have been modified to remove all references to Table
3.6-1. This change is in accordance with the guidance provided in
Generic Letter 91-08, ``Removal of Component Lists from Technical
Specifications,'' dated May 6, 1991.
Date of issuance: July 23, 1997
Effective date: July 23, 1997, to be implemented within 30 days
from the date of issuance.
Amendment No.: 108
Facility Operating License No. NPF-42: The amendment revised the
Technical Specifications and the Operating License.
Date of initial notice in Federal Register: April 23, 1997 (62 FR
19838) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 23, 1997. 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 6th day of August, 1997.
For the Nuclear Regulatory Commission
Jack W. Roe,
Director, Division of Reactor Projects III/IV, Office of Nuclear
Reactor Regulation.
[Doc. 97-21244 Filed 8-12-97; 8:45 am]
BILLING CODE 7590-01-F
