[Federal Register Volume 59, Number 32 (Wednesday, February 16, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X94-10216]
[[Page Unknown]]
[Federal Register: February 16, 1994]
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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 January 22, 1994, through February 4, 1994.
The last biweekly notice was published on February 2, 1994 (59 FR
4933).Notice Of Consideration Of Issuance Of Amendments To Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, And Opportunity For A Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, and should cite the publication date and page
number of this Federal Register notice. Written comments may also be
delivered to Room P-223, Phillips Building, 7920 Norfolk Avenue,
Bethesda, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies
of written comments received may be examined at the NRC Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC 20555.
The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By March 18, 1994, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC 20555 and at the local
public document room for the particular facility involved. If a request
for a hearing or petition for leave to intervene is filed by the above
date, the Commission or an Atomic Safety and Licensing Board,
designated by the Commission or by the Chairman of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the designated Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington DC 20555, by the
above date. Where petitions are filed during the last 10 days of the
notice period, it is requested that the petitioner promptly so inform
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and
to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC 20555, and at the local public document
room for the particular facility involved.
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: January 24, 1994
Description of amendment request: The proposed amendment would
implement Line Item 5.9 of NRC Generic Letter 93-05, ``Line Item
Technical Specification Improvements to Reduce Surveillance
Requirements for Testing During Power Operation,'' which recommends
licensees consider deleting the requirements to perform response time
testing for selected instrumentation in the isolation system where the
required time corresponds to the diesel start time.
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:
LaSalle has evaluated the proposed Technical Specification
Amendment. Based upon the criteria for defining a Significant
Hazards Consideration established in 10 CFR 50.92(c), operation of
LaSalle County Station in accordance with the proposed amendment
will not:
1) Involve a significant increase in the probability or
consequences of an accident previously evaluated because:
The proposal seeks to eliminate response time testing
requirements for selected instrumentation in the isolation system.
The proposal does not introduce changes in the response times
themselves. The probability and consequences of an accident
previously evaluated are not increased because accepted licensing
criteria are maintained. The requirements for channel checks,
functional tests, calibrations, and logic system functional tests
are not altered by this proposal. The ability to detect degrading
trends of response times is available via the above Technical
Specification required tests. Therefore, the response times of these
systems will be maintained within the acceptance limits assumed in
plant safety analyses and required for successful mitigation of an
initiating event because of the continued Technical Specification
testing.
2) Create the possibility of a new or different kind of accident
from any accident previously evaluated because:
The proposal does not change component or system interactions.
Accident analyses assume a loss of AC power which is restored by
startup of emergency diesel generators. The 13 second interval
associated with the restoration of AC power, which establishes the
response time for the isolation functions, is maintained. The
starting, sequencing, and loading functions associated with the
diesel generators is not affected by the proposed change. The
response times include the instrument response times, which are
typically measured in fractions of a second, and the response times
of the actuation logic circuits, which are typically less than a
second. These times are small in comparison to the diesel generator
start time (13 seconds). The ability of the isolation system to
perform its intended function to mitigate the consequences of an
initiating event within the acceptance limits assumed in plant
safety analyses is not altered by the proposed change.
3) Involve a significant reduction in a margin of safety
because:
The proposal does not involve the relaxation of any criteria
identified in the SAR or reduce any of the requirements of Technical
Specifications. The proposed revision does not affect licensing
acceptance limits associated with accidents. With the exception of
MSIVs, the safety analyses do not address individual sensor response
times or the response times of the logic systems to which the
sensors are connected. These analyses conservatively establish the
margin of safety. Deleting the requirement to perform unnecessary
response time testing does not affect the results of accident and
transient analyses. Plant and system response to an initiating event
will remain in compliance within the assumptions of safety analyses.
The proposed change does not increase the probability or
consequences of an accident, and there is no impact on equipment
important to safety or systems, structures or components. There is
no associated change to the type, amount, or control of radioactive
effluents, nor is there an associated increase in individual or
cumulative occupational radiation exposure. There is no effect upon
the capabilities of the associated systems to perform their intended
functions within the allowed response times assumed in safety
analyses. Therefore, the margin of safety is preserved.
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: Public Library of Illinois
Valley Community College, Rural Route No. 1, Ogelsby, Illinois 61348
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690
NRC Project Director: James E. Dyer
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam
Neck Plant, Middlesex County, Connecticut
Date of amendment request: January 17, 1994
Description of amendment request: Connecticut Yankee Atomic Power
Company (CYAPCO) proposes to remove Technical Specification 3/4.4.12,
``Failed Fuel Rods'' and its associated BASES Section 3/4.4.12.
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 (SHC), which is presented below:
The proposed changes do not involve an SHC consideration because
the changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
A review of the accidents detailed in the Updated Final Safety
Analysis Report, Chapter 15, was undertaken to determine if they
were impacted by the proposed change. The review indicated that the
previously evaluated accidents were not impacted by the proposed
license amendment.
All fuel design and performance criteria are the same for Cycle
18 as in previous cycles. All criteria will continue to be met and
no new single-failure mechanisms will be created. This change does
not involve any alterations to plant equipment or procedures which
would affect any operational modes or accident assumptions. This
proposed license amendment does delete a technical specification
that is no longer considered necessary. This deletion is prompted by
the replacement of stainless steel clad fuel with zircaloy clad
fuel. The zircaloy clad fuel, if it experiences damage, will release
iodine into the primary system. Any iodine released is covered
within the guidelines specified in the existing Technical
Specification 3/4.4.8, ``Specific Activity.'' This specification
will ensure that operation does not continue with radiochemistry
values that exceed those assumed in our accident assumptions. The
existing Technical Specification of specific activity along with the
zircaloy clad fuel will ensure that a significant increase in the
probability or consequences of an accident previously evaluated is
not present.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The possibility of an accident or malfunction of a different
type than any evaluated previously in the UFSAR [Updated Final
Safety Analysis Report] is not created. Since there are no changes
in the way the plant is operated, the potential for an unanalyzed
accident is not created. No new failure modes are introduced.
The presence of defective fuel rods and the resultant iodine
release would only affect potential offsite doses. This proposed
license amendment does not increase the radiochemistry limits, but
does revert the technical specifications back to the standard
methodology and limitations that were unable to be used because of
the stainless steel clad fuel. These new limitations will continue
to ensure that doses remain within the limits prescribed.
3. Involve a significant reduction in a margin of safety.
The proposed changes do not have any adverse impact on the
protective boundaries. The margin of safety, as defined in the basis
for any technical specification, is not reduced. The proposed
changes do not adversely impact any of the safety systems, nor do
they increase the number of challenges to the safety systems.
The limit of 160 defective rods was chosen to be consistent with
initial conditions assumed for the radiological design basis. The
elimination of this specification is acceptable since the basis for
the initial condition can be supported by the use of zircaloy clad
fuel as opposed to the unique stainless steel clad. If future fuel
defects are debris induced, the dose equivalent iodine will be
within expected radiochemistry values and the resulting doses will
be bounded. Therefore, there is no reduction in the margin of safety
as defined in the basis of any technical specification with the
deletion of the defective fuel rod technical specification.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Russell Library, 123 Broad
Street, Middletown, Connecticut 06457.
Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &
Howard, Counselors at Law, City Place, Hartford, Connecticut 06103-
3499.
NRC Project Director: John F. Stolz
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of amendment request: December 6, 1993
Description of amendment request: The proposed amendment request
would revise the Technical Specifications (TSs) to provide a temporary
one-time revision to the Definition Section of the TS. Specifically, a
footnote is added in the Definition Section of the TS which is
applicable to TS 1.2.1, ``Cold Shutdown Condition,'' changing Tavq
less than or equal to 200 deg.F to less than or equal to 250 deg.F and
TS 1.2.2, ``Hot Shutdown Condition,'' changing Tavq greater than
200 deg.F to greater than 250 deg.F. The footnote further states that
the change is for the one time, fuel out, chemical decontamination
program. This program is currently scheduled for the upcoming 1995
refueling outage of the Indian Point Nuclear Generating Unit 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed change does not involve a significant hazards
consideration since:
1. There is no significant increase in the probability or
consequences of an accident.
Approval of the proposed one time change to the Technical
Specification definition of cold shutdown for purposes of performing
the full RCS [reactor coolant system] chemical decontamination
without fuel in the reactor would provide relief from unnecessary
technical specification action statements that are based on fuel in
the reactor. Credible accidents with significant consequences are
practically eliminated with the removal of the reactor fuel during
the performance of the FSD [full reactor coolant system chemical
decontamination]. In addition, specific actions would be taken in
accordance with the requirements of the NRC approved WCAP-12932-A
Rev. 2 to ensure that RCS and affected interfacing systems integrity
are preserved. Thus, system capability within established accident
scenarios would not be compromised. The proposed amendment would
therefore not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The possibility of a new or different kind of accident from
any previously analyzed has not been created.
As noted above, the proposed amendment seeks to eliminate
unnecessary Technical Specification action requirements during the
performance of full RCS chemical decontamination. These actions are
unnecessary because there will be no fuel in the reactor and the RCS
and other affected systems will be operated under conditions well
within their design capability during the implementation of this
process. In addition, the FSD effort will be conducted in accordance
with the requirement(s) of the NRC approved Westinghouse topical
report WCAP-12932-A Rev. 2. Accidents involving failures of the
decontamination process system will not exceed the bounding
conditions for any previously established accidents involving
failure of a radwaste system. Accordingly, the possibility of a new
or different kind of accident from any previously analyzed will not
be created.
3. There has been no reduction in the margin of safety.
The proposed amendment provides relief from technical
specification actions in the performance of the FSD which become
unnecessary when there is no fuel in the reactor. The change will
not adversely impact any Technical Specification required systems,
structures or components. The design capability of systems,
structures or components impacted will not be reduced. Consequently,
no significant reduction in the margin of safety for any system,
structure, or component is involved.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,
New York, New York 10003.
NRC Project Director: Robert A. Capra
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 amendment request: January 18, 1994
Description of amendment request: The proposed amendments remove
the tables of containment penetration conductor overcurrent protective
devices from the Technical Specifications (TS) in accordance with the
guidance contained in Generic Letter 91-08, ``Removal of Component
Lists from Technical Specifications.'' The tables would be relocated to
Chapter 16 of the Catawba Final Safety Analysis Report (Selected
Licensee Commitments Manual). In addition, the licensee proposes the
removal of an obsolete footnote to TS 4.8.4. The footnote, which made
TS 4.8.4.a initially effective following the first refueling outage of
Unit 1, is no longer needed.
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 requested amendments will not involve a significant increase
in the probability or consequences of an accident previously
evaluated. Relocating the component lists of containment penetration
conductor overcurrent protective devices from the technical
specifications to the [Selected Licensee Commitments] SLC Manual
(with all attendant required technical specification changes as
described previously and also including removal of the above
described obsolete footnote) has no impact upon either the
probability or consequences of any accident. No plant equipment is
affected by the proposed change. No equipment is being added or
deleted from the lists; only the source document for the lists is
being changed. Any future changes to the lists (i.e., changes to the
plant) will be subject to the provisions of 10CFR50.59 and also
subject to the change control provisions of Chapter 6 of Catawba's
Technical Specifications.
Criterion 2
The requested amendments will not create the possibility of a
new or different kind of accident from any accident previously
evaluated. No accident causal mechanisms are affected by the
proposed change, as no change to the plant is being proposed. In
addition, no change to the manner in which the plant is operated is
being made. Finally, no changes to plant procedures are being made
which would affect any accident causal mechanisms.
Criterion 3
The requested amendments will not involve a significant
reduction in a margin of safety. The proposed change has no impact
upon any safety margin. The proposed change is consistent with the
guidance provided in Generic Letter 91-08 and the control provisions
utilized as a result of relocating the subject component lists are
at least as stringent as those set forth in the generic letter.
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: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242
NRC Project Director: Loren R. Plisco, Acting
Duke Power Company, Docket No. 50-413, Catawba Nuclear Station,
Unit No. 1, York County, South Carolina
Date of amendment request: January 10, 1994
Description of amendment request: The proposed amendment would
revise Technical Specifications 2.0 and 3/4.2 which currently requires
the determination of the reactor coolant system flow rate by precision
heat balance measurement at least once per 18 months. Date of
publication of individual notice in Federal Register: January 26, 1994
(59 FR 3743)
Expiration date of individual notice: February 25, 1994
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: January 10, 1994
Description of amendment request: The amendments would change the
method of measuring the reactor coolant system flow rate (Technical
Specifications 2.0 and 3/4.2) during the 18-month surveillance for
McGuire, Units 1 and 2.
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) This amendment will not significantly increase the probability
or consequence of any accident previously evaluated.
No component modification, system realignment, or change in
operating procedure will occur which could affect the probability of
any accident or transient. The change in method of flow measurement
will not change the probability of actuation of any Engineered
Safeguard Feature or other device. The actual flow rate will not
change. The consequences of previously-analyzed accidents will not
change as a result of the new method of flow measurement.
(2) This amendment will not create the possibility of any new or
different accidents not previously evaluated.
No component modification or system realignment will occur which
could create the possibility of a new event not previously
considered. The elbow taps are already in place, and are used to
monitor flow for the Reactor Protection System. They will not
initiate any new events.
(3) This amendment will not involve a significant reduction in a
margin of safety.
As described in [the licensee's application], the change in
method of RCS flow measurement will provide a more accurate
indication of the flow. The actual flow rate will not be affected.
The revised setpoints for low reactor coolant flow are driven by
changes to statistical allowances and do not represent substantive,
or less conservative, changes. There is no significant reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Astkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242
NRC Project Director: Loren R. Plisco, Acting
Entergy Operations Inc., Docket No. 50-382, Waterford Steam
ElectricStation, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: December 23, 1993
Description of amendment request: The proposed amendment would
revise the Technical Specification (TS) for the following four items in
accordance with the guidance in Generic Letter (GL) 93-05 ``Line Item
Technical Specifications Improvements To Reduce Surveillance
Requirements For Testing During Power Operation''.1) GL Item 5.14
Radiation Monitors will change the channel functional test from monthly
to quarterly.2) GL Item 6.1 Reactor Coolant System (RCS) Isolation
Valves will increase the time from 72 hours to 7 days for remaining in
cold shutdown without leak testing the RCS isolation valves.3) GL Item
6.6 Pressurizer Heaters will change the verification of capacity from
at least once per 92 days to each refueling outage and will change the
demonstration of the emergency power supply from at least once per 18
months to at each refueling outage.4) GL Item 9.1 Auxiliary Feedwater
Pump and System Testing will change the frequency of these pumps from
once per 31 days on a staggered basis to quarterly on a staggered
bases.
All of the above are compatible with Waterford 3 plant operating
experience and are consistent with NUREG-1366, ``Improvement To
Technical Specification Surveillance Requirements,'' December 1992 and
the licensing basis for Waterford 3.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed change to increase the radiation monitoring
instrumentation channel functional test from monthly to quarterly
will have no effect on design basis accidents. The findings in
NUREG-1366 determined that this change will increase the
availability of radiation monitors.
The proposed change to increase the 72 hour time for remaining
in cold shutdown without leak testing the RCS isolation valves to 7
days will not affect any design basis accidents. NUREG-1366 findings
have determined that extending this interval does not significantly
alter the associated risk. In addition, the current requirement has
a potential for causing problems resulting from a hurried recovery.
The proposed change to the pressurizer heater capacity test
interval from quarterly to each refueling interval will have no
affect on any design basis accidents. The TS requires at least 2
groups of pressurizer heaters each having a nominal capacity of 150
kW. Waterford 3 has 8 groups of pressurizer heaters; two
proportional groups of 150 kW each, and 6 backup groups of 200 kW
each. An evaluation of past operating experience has shown the
availability of at least 6 groups of pressurizer heaters with a
minimum of 150 kW each.
The proposed change to extend the testing interval for the EFW
[emergency feedwater] pumps will have no affect on any design basis
accidents. The pumps will continue to be tested quarterly to the
same standards applied to safety related pumps as defined by the
ASME [American Society of Mechanical Engineers] Section XI Code.
Satisfactory completion of testing in accordance with the Code is
accepted as verification that safety related pumps will be available
to perform their intended function.
The proposed changes identified above are supported by the
findings identified in NUREG-1366 and consistent with the guidance
provided in Generic Letter 93-05. These line-item improvements are
intended to improve plant safety, decrease equipment degradation,
and remove unnecessary burden on personnel resources by reducing the
amount of testing that the TS require during power operation.
Therefore, the proposed changes identified above will not involve a
significant increase in the probability or consequences of any
accident previously evaluated.
The changes identified above only affect the frequency of
surveillance testing. There are no changes that will alter operation
of the plant or the manner in which it is operated. Therefore, the
proposed changes will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes identified herein extend testing frequency
in an effort to improve plant reliability and safety. The proposed
changes are consistent with the findings in NUREG-1366, guidance in
Generic Letter 93-05 and plant operating experience. As such, the
proposed changes will preserve the established margin of safety for
the affected specifications. Therefore, the proposed changes will
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122
Attorney for licensee: N. S. Reynolds, Esq., Winston & Strawn 1400
L Street NW., Washington, D.C. 20005-3502
NRC Project Director: William D. Beckner
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2,
Burke County, Georgia
Date of amendment request: November 19, 1993
Description of amendment request: The proposed change would
relocate the requirements of Technical Specification 3/4.3.4, Turbine
Overspeed Protection, to Section 16.3 of the Vogtle Electric Generating
Plant, Units 1 and 2, 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 change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The proposed change involves the relocation of the TS
[Technical Specification] requirements for the turbine overspeed
protection system to the VEGP [Vogtle Electric Generating Plant]
FSAR [Final Safety Analysis Report]. The requirements that will
reside in the FSAR will continue to ensure that the probability of
turbine missile generation is maintained below NRC limits as defined
in NUREG-1048, Appendix U. Since the turbine overspeed protection
system will remain capable of protecting the turbine from excessive
overspeed, the proposed change will have no effect on the
consequences of an accident previously evaluated.
2. The proposed change will not create the possibility of a new
or different kind of accident than any previously evaluated. The
proposed change does not involve any change to the configuration or
method of operation of any plant equipment, and no new failure modes
have been defined for any plant system or component. In addition, no
new limiting failures have been identified as a result of the
proposed change. The requirements for the turbine overspeed
protection system that will reside in the FSAR will ensure that the
system remains capable of protecting the turbine from excessive
overspeed. Therefore, the proposed change will not create the
possibility of a new or different kind of accident than any
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety. The proposed change would allow the
requirements for the turbine overspeed protection system to be
relocated to the FSAR on the basis that the turbine overspeed
protection system does not meet the criteria of the NRC Final Policy
Statement on Technical Specifications Improvements for Nuclear
Reactors. The requirements that will reside in the FSAR for the
turbine overspeed protection system will ensure that the system
remains capable of protecting the turbine from excessive overspeed.
Therefore, the proposed change 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: Burke County Public Library,
412 Fourth Street, Waynesboro, Georgia 30830.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308
NRC Project Director: Loren R. Plisco, Acting
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket No. 50-499 South Texas Project, Unit 2, Matagorda County,
Texas
Date of amendment request: January 25, 1994
Description of amendment request: The licensee proposes to make a
one-time change to the technical specifications to add new Technical
Specifications 3/4.10.6 and 3/4.10.7 to the Special Test Exemptions
section. The new TS would allow the restart of Unit 2 with expired
calibrations on the core exit thermocouples (CET) and the reactor
coolant system (RCS) resistance temperature detectors (RTD). This
amendment will also add a new Technical Specification to allow the
ascension to 75 percent rated thermal power with an expired precision
heat balance reactor coolant flow measurement.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1)The proposed change does not involve a significant increase
in the probability or consequences of a previously evaluated
accident.
The proposed change will allow the restart of [STP] Unit 2 with
Core Exit Thermocouples and Reactor Coolant System Resistance
Temperature Detectors technically inoperable due to expired
calibrations. The calibrations of these instruments can only be
completed when the Unit reaches Normal Operating Pressure and Normal
Operating Temperature in Mode 3. Once the calibrations of these
instruments are completed, this one time change will expire and all
of the existing applicable Limiting Conditions for Operations will
become effective immediately. Since industry and South Texas Project
Electric Generating Station experience has shown that the failure
mechanism for these types of instrument is complete failure as
opposed to a gradual drift, and there will be calibration points to
compare RTD readings to actual RCS temperature as the RCS
temperature increases, it is reasonable to expect these CETs/RTDs
will function as they did before their calibrations expired. For
this reason, all applicable functions, including COMS, Thot ,
Tcold, and Tavg are expected to operate normally. Because
normal operation of the instruments is expected and the only reason
for the instruments being declared inoperable is their expired
calibrations, this change does not involve a significant increase in
the probability or consequence of an accident previously evaluated.
The proposed change will also allow the restart of Unit 2 with
the precision heat balance RCS flow measurement surveillance
expired. This surveillance is used to confirm the values indicated
by the RCS flow meters. These instruments are calibrated every 18
months and the RCS flow meters will be checked every 12 hours to
ensure adequate flow prior to the completion of the precision heat
balance RCS flow measurement. Since this surveillance is only used
to confirm the reading of calibrated instruments and does not
involve any changes to the design or function of the instruments,
this change does not involve a significant increase in the
probability or consequence of an accident previously evaluated.
(2) The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated.
The operations of Unit 2 with the CETs and RCS RTDs technically
inoperable due to expired calibrations, until these calibrations can
be completed in Mode 3, does not affect the design bases of the CETs
and RCS RTDs or any of the accident evaluations involving these
instruments. Since industry and South Texas Project Electric
Generating Station experience indicates that the failure mechanism
for these types of instruments is not a gradual drift but complete
failure, the reasonable expectation is the CETs/RTDs will function
as they did prior to their calibrations expiring.
Additionally, the operation of Unit 2 with the precision heat
balance RCS flow measurement surveillance expired does not affect
the design bases of the RCS flow meters or any of the accident
evaluations involving these instruments. This surveillance is used
to confirm the values indicated by the RCS flow meters. These
instruments are calibrated every 18 months and the RCS flow meters
will be checked every 12 hours to ensure adequate flow prior to the
completion of the precision heat balance RCS flow measurement.
Because normal operation of all of these instruments is
expected, these changes do 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 the margin of safety.
The RCS RTDs are auctioneered to prevent a failed high or low
instrument from adversely influencing the safety of the plant. This
feature is still operable and will, along with normal operator
activities, provide assurance that the margin of safety is not
reduced by this change. In addition, the change does not affect the
design bases, accident analysis, reliability or capability of the
CETs/RTDs to perform their intended safety functions. The RCS flow
meters will be checked every 12 hours to ensure adequate flow prior
to the completion of the precision heat balance RCS flow
measurement.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendment involves no significant hazards
consideration.Local Public Document Location: Wharton County Junior
College, J.M. Hodges Learning Center, 911 Boling Highway, Wharton,
Texas 77488
Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger,
P.C., 1615 L Street, NW, Washington, DC 20036
NRC Project Director: Suzanne C. Black
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: September 28, 1993
Description of amendment request: The proposed amendment would
revise the Cooper Nuclear Station (CNS) Technical Specifications to
modify the licensee's organizational structure by removing the
positions of ``Site Manager'' and ``Senior Manager of Operations.'' The
functions presently given in CNS Technical Specifications for the Site
Manager position will be assumed by the Vice President - Nuclear.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Evaluation
The proposed change removing the positions of Site Manager and
Senior Manager of Operations from the Technical Specifications is
administrative in nature. The functions and responsibilities of the
previous position of Site Manager presently given in the plant
Technical Specifications will be performed by the Vice President -
Nuclear. Additionally, with the reorganization, the Senior Manager
of Operations position is eliminated and therefore, this position is
also being removed. The provision in the Technical Specifications
for automatic shifting of Plant Manager responsibilities to the
Senior Manager of Operations has also been removed. The shifting of
Plant Manager responsibilities (in writing) to one of the Managers
at CNS who is qualified for this position remains in the Technical
Specifications. The position removals and responsibility transfers
in the organization do not affect plant design or operation, nor do
they affect the way any systems, structures, or components are
operated or maintained. The individual filling the position ``Vice
President - Nuclear'' is qualified to perform the assigned tasks and
responsibilities. Restructuring of the sentence in specification
6.2.B.6, is purely an administrative change. Also, this proposed
change does not alter the conditions or assumptions in any of the
Updated Safety Analysis Report (USAR) accident analyses. Since the
USAR accident analyses remain bounding, the consequences previously
evaluated are not adversely affected by the proposed change.
Therefore, it can be concluded that the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed License Amendment create the possibility of
a new or different kind of accident from any accident previously
evaluated?
Evaluation
The proposed Technical Specification revision removes all
references to the position title of the Site Manager. The
responsibilities of this position presently given in the Technical
Specifications are being incorporated and performed by the position
``Vice President - Nuclear.'' Additionally, with the reorganization,
the Senior Manager of Operations position is eliminated and
therefore, this position is also being removed. The shifting of
Plant Manager responsibilities (in writing) to one of the Managers
at CNS who is qualified for this position remains in the Technical
Specifications. All given management activities will continue to be
performed by qualified individuals. Restructuring of the sentence in
specification 6.2.B.6 is purely an administrative change. This
change does not affect the design or operation of any system,
structure, or component in the plant, and is considered to be an
administrative change. Accordingly, no new failure modes have been
defined for any plant system or component important to safety, nor
has any new limiting failure been identified as a result of the
proposed change. Therefore, this proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in the margin of safety?
Evaluation
This proposed amendment involves a change to the Administrative
Controls Section of the CNS Technical Specifications; specifically,
removal of two positions referenced in the organizational structure.
The Site Manager position is being deleted and the responsibilities
of this position listed in the Technical Specifications are being
performed by the Vice President - Nuclear. Additionally, with the
reorganization, the Senior Manager of Operations position is
eliminated and therefore, this position and responsibilities are
also being removed. The shifting of Plant Manager responsibilities
(in writing) to one of the Managers at CNS who is qualified for this
position remains in the Technical Specifications. All given
management activities, as described in the Technical Specifications,
will continue to be performed by qualified individuals.
Restructuring of the sentence in specification 6.2.B.6, is purely an
administrative change. The proposed change does not adversely impact
the plant's ability to meet applicable regulatory requirements. The
proposed change does not alter any means of plant operation, nor
does the proposed change involve any physical alterations to the
plant and does not affect any plant safety parameters or setpoints.
Therefore, this proposed change 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 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: Asuburn Public Library, 118
15th Street, Auburn, Nebraska 68305
Attorney for licensee: Mr. G. D. Watson, Nebraska Public Power
District, Post Office Box 499, Columbus, Nebraska 68602-0499
NRC Project Director: William D. Beckner
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: December 10, 1993
Description of amendment request: The proposed amendment would
revise the Cooper Nuclear Station (CNS) Technical Specifications
Sections 3/4.21 ``Environmental/Radiological Effluents,'' and 6.5,
``Station Reporting Requirements,'' to change the frequency of the
reporting period of the ``Semiannual Radioactive Materials Release
Report'' from semiannual to annual and to extend the reporting
frequency of the Annual Design Change Report from an annual submittal
to annually or along with the Updated Safety Analysis Report (USAR)
updates required by 10 CFR 50.71(e). These proposed changes are
intended to make the CNS Technical Specifications consistent with the
current provisions of 10 CFR 50.36(a) and 10 CFR 50.59(b),
respectively.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Evaluation
The proposed changes are administrative in nature and makes the
Cooper Nuclear Station (CNS) Technical Specifications (T/S)
consistent with amended regulations of 10CFR50.36(a), and 10CFR
50.59(b) by reducing the submittal frequency of certain reports to
the NRC. The proposed revisions do not involve any change to plant
design, plant operation, or configuration of any plant equipment
that is used to mitigate the consequences of an accident previously
evaluated. Also, the proposed changes do not alter the conditions or
assumptions in any of the Updated Safety Analysis Report (USAR)
accident analyses. Since the USAR accident analyses remain bounding,
the radiological consequences previously evaluated are not adversely
affected by the proposed changes. As administrative changes, all
defined terms on the affected pages have been capitalized.
Therefore, it can be concluded that the proposed changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed License Amendment create the possibility of
a new or different kind of accident from any accident previously
evaluated?
Evaluation
The proposed changes are administrative in nature and makes the
CNS T/S consistent with amended regulations of 10CFR50.36(a), and
10CFR50.59(b) by reducing the submittal frequency of certain reports
to the NRC. The proposed revisions do not involve any change to
plant design, plant operation, or configuration of any plant
equipment that is used to mitigate the consequences of an accident
previously evaluated. Accordingly, no new failure modes have been
created for any plant system or component important to safety nor
has any new limiting failure been identified as a result of the
proposed changes. Also, there will be no change in the types or
increase in the amount of effluents released offsite. As
administrative changes, all defined terms on the affected pages have
been capitalized. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change create a significant reduction in
the margin of safety?
Evaluation
The proposed changes are administrative in nature and do not
adversely impact the plant's ability to meet applicable regulatory
requirements related to liquid or gaseous effluents, and solid waste
releases. The proposed changes do not alter any administrative
controls over radioactive effluents, nor do the proposed changes
involve any physical alterations to the plant with respect to
radioactive effluents. These changes do not affect the meaning,
application, and function of the T/S requirements. The proposed
change will reduce the administrative burden of NRC reporting
without reducing the protection for public health and safety. As
administrative changes, all defined terms on the affected pages have
been capitalized. Therefore, the proposed change 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 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: Asuburn Public Library, 118
15th Street, Auburn, Nebraska 68305
Attorney for licensee: Mr. G. D. Watson, Nebraska Public Power
District, Post Office Box 499, Columbus, Nebraska 68602-0499
NRC Project Director: William D. Beckner
Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile
Point Nuclear Station Unit No. 1, Oswego County, New York
Date of amendment request: January 6, 1994
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Tables 3.2.7, 3.6.2a, 4.6.2a,
3.6.2b and 4.6.2b to delete the main steam line isolation and automatic
reactor shutdown (reactor scram) functions of the Main Steam Line
Radiation Monitor. Conforming changes would also be made to the Bases
of these TSs and to the Bases for TS 2.1.2. The licensee stated that
the proposed changes would be consistent with the NRC's Improved
Standard Technical Specifications, NUREG-1433, and with NRC-approved
(Safety Evaluation, dated May 15, 1991) Boiling Water Reactor Owners'
Group Licensing Topical Report NEDO-31400A, dated July 9, 1987.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The operation of Nine Mile Point Unit 1 in accordance with the
proposed amendment will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated
because the functions being removed do not contribute to avoidance
of any previously evaluated accidents. Further, the changes have
been shown to have an insignificant impact on overall reactivity
control failure frequency. This insignificant impact is offset by
the relatively large reduction in core damage frequency realized by
the implementation of these changes. Hence, the probability or
consequences of previously evaluated accidents are not significantly
increased due to this change. To the contrary, as stated in the
topical report [NEDO-31400A] the changes provide a net improvement
in overall plant safety.
The proposed amendment does not involve a physical or procedural
change to any structure, component or system that significantly
affects the probability or consequences of any accident or
malfunction of equipment important to safety previously evaluated in
the Final Safety Analysis Report (Updated). The proposed amendment
will involve a change to reactor protection and isolation actuation
systems circuitry that will remove the automatic reactor shutdown
and Main Steam Line Isolation Valve closure functions of the Main
Steam Line Radiation Monitor. However, the physical changes will not
affect the remaining scram or vessel isolation functions.
[***T]he methods, procedures and assumptions used to perform the
eneric analyses in NED0-31400A are bounding for the Nine Mile Point
Unit 1 with regard to input values. Niagara Mohawk has also provided
in the evaluation reasonable assurance that significantly increased
levels of radioactivity in the main steam lines will be controlled
expeditiously to limit both occupational and environmental
exposures. The Main Steam Line Radiation Monitor alarm setpoints
will be set at 1.5 times the normal full power background dose rate
and should any monitor exceed its alarm setpoint, the reactor
coolant will be sampled to determine activity levels and the
possible need for additional corrective actions.
The offgas radiation monitor is a more sensitive monitor than
the Main Steam Line Radiation Monitor because the nitrogen-16
source, dominating the radiation signal to the Main Steam Line
Radiation Monitor, has decayed by the time the radiation monitor can
be affected by any increased levels of activity. Therefore, setting
the offgas radiation monitor at 1.5 times the nitrogen-16 background
dose rate is not reasonable since setting the monitor that low can
lead to spurious activations of the alarm.
Nine Mile Point Unit 1's monitor configuration, as described in
the FSAR, detects the concentration of the offgas as it flows
through the pipe. Thus, the detector is sensitive to fluctuations in
condenser air inleakage, which can have an appreciable impact on the
monitor readings, especially at readings as low as 1.5 times the
normal full power background. Therefore, Niagara Mohawk proposes to
set the alarm at five (5) times the normal full power background,
which is still very conservative compared to the value allowed by
Technical Specification 3.6.15.c., which is set based on Nine Mile
Point Unit 1's Offsite Dose Calculation Manual.
Niagara Mohawk believes that a setting of five (5) times the
normal full power background is extremely conservative and is low
enough to ensure detection of even minor fuel performance changes.
Furthermore, if the monitor alarms at this setpoint of five times
the normal full power background, the offgas will immediately be
sampled and analyzed, followed by an analysis of a reactor coolant
sample.
Furthermore, the analyses in the Licensing Topical Report
demonstrate that removal of the automatic reactor scram and Main
Steam Line Isolation Valve closure functions of the Main Steam Line
Radiation Monitor does not change the conclusions in the Final
Safety Analysis Report (Updated) that the calculated radiological
release consequences of the bounding control rod drop accident will
not exceed the acceptable dose limits specified in 10CFR[Part]100.
Therefore, Niagara Mohawk concludes that the proposed amendment
will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The operation of Nine Mile Point Unit 1 in accordance with the
proposed amendment will not create the possibility of a new or
different kind of accident from any previously evaluated.
The function of a Main Steam Line Radiation Monitor trip is to
detect abnormal fission produce release and isolate the main steam
lines, thereby stopping the transport of fission products from the
reactor to the main condenser. The monitors do not perform a
prevention function for any kind of accident.
The main steam line high radiation scram and main steam line
isolation functions were originally intended to mitigate, not
prevent, an existing accident scenario. However, the functions being
removed do not contribute to avoidance or mitigation of any
previously evaluated accidents since no credit is taken for these
functions in any design basis event for terminating the initiating
event or assuring the radioactive release remains within accepted
limits. The existence of a Main Steam Line Radiation Monitor trip
does not prevent the occurrence of a fuel failure event or any other
type of event. Elimination of these functions will not introduce a
new or different accident scenario.
The proposed amendment represents a change to the physical
configuration of the plant in that some reactor protection system
circuits will be modified to eliminate the main steam line high
radiation scram and main steam line isolation signals. However,
these changes will not affect the remaining scram or vessel
isolation functions. In all other respects, plant design and
operation remain unchanged.
Therefore, Niagara Mohawk Power Corporation concludes that the
proposed amendment will not create the possibility of a new or
different kind of accident from any previously evaluated.
The operation of Nine Mile Point Unit 1 in accordance with the
proposed amendment will not involve a significant reduction in a
margin of safety.
The proposed changes do not involve a significant reduction in a
margin of safety because, as shown in the topical report, the
changes represent an overall improvement in plant safety in that the
core damage frequency is reduced. Safe operation of the plant is
enhanced by elimination of the unnecessary scram and isolation of
the reactor vessel. With implementation of these changes, the
primary heat sink remains available, a large transient on the vessel
and safety-related actuations is avoided, and the Offgas System
remains available to control the pathway of a potential release.
Therefore, Niagara Mohawk concludes that 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: Robert A. Capra
Northeast Nuclear Energy Company, Docket No. 50-245, Millstone
NuclearPower Station, Unit 1, New London County, Connecticut
Date of amendment request: January 14, 1994
Description of amendment request: The proposed amendment corrects
an editorial error. Specifically, the amendment changes the reference
in Limiting Condition for Operation (LCO) 3.4.D from ``3.3.A through
C'' to ``3.4.A, 3.4.B, and 3.4.C.'' The amendment also changes the
associated bases to clarify the LCO minimum solution concentration
requirement of 11 weight percent and updates the excerpt from 10 CFR
50.62 to reflect the current text of the regulation.
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:
In accordance with 10 CFR 50.92, NNECO [Northeast Nuclear Energy
Company] has reviewed the proposed change and has concluded that it
does not involve a significant hazards consideration (SHC). The
basis for this conclusion is that the three criteria of 10 CFR
50.92(c) are not compromised. The proposed change does not involve
an SHC because the change would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change only modifies an incorrect reference in
Section 3.4.D of the Technical Specifications. In practice, if
Specification 3.4.A, 3.4.B, or 3.4.C cannot be met, an orderly
shutdown is initiated. As currently written, the failure to meet the
requirements of Section 3.3 would also initiate a shutdown in
accordance with Section 3.4.D. This is not the intent of Section
3.4.D since Section 3.3 already has specific shutdown requirements.
This proposed change will correct Section 3.4.D so that it limits
the conditions under which a plant shutdown must be initiated to the
LCOs of the standby liquid control system. Therefore, this proposed
change will not increase the probability or consequences of an
accident.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
The proposed change corrects an incorrect section reference.
There is no change to the operation or design of the plant, nor is
there any change to the operability requirements of either section.
The proposed change properly identifies the conditions under which
the plant must be shutdown if an LCO is not met for the standby
liquid control system. In practice, if Specification 3.4.A, 3.4.B,
or 3.4.C cannot be met, an orderly shutdown is initiated. Since
there is no change in plant operation or design, there is no
possibility of a different kind of accident.
3. Involve a significant reduction in a margin of safety.
The proposed change does not modify the design or function of
the plant, nor does it reduce operability requirements of either
Section 3.3 or 3.4. The proposed change only corrects an incorrect
section reference by identifying the correct shutdown requirements
for the standby liquid control system. Since there is no change to
plant operation or design and the shutdown requirements are not
reduced, there is no reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, Connecticut 06360.
Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &
Howard, Counselors at Law, City Place, Hartford, Connecticut 06103-
3499.
NRC Project Director: John F. Stolz
Northeast Nuclear Energy Company (NNECO), Docket Nos. 50-245, 50-
336 and 50-423, Millstone Nuclear Power Station, Units 1, 2 and 3,
New London County, Connecticut
Date of amendment request: December 22, 1993
Description of amendment request: The proposed amendments would
change the Technical Specification (TS) as follows:
1. Change the title of the Nuclear Station Director to Senior Vice
President - Millstone Station.
2. Remove the requirement to provide a copy of Plant Operations
Review Committee (PORC) and Site Operations Review Committee (SORC)
meeting minutes to the Executive Vice President - Nuclear. The Senior
Vice President - Millstone Station is being proposed to replace the
Executive Vice President - Nuclear for receipt of PORC and SORC meeting
minutes.
3. Make editorial changes to the Millstone Unit No. 1 TS Index.
4. Correct a typographical error in Section 6.2.1.d of the
Millstone Unit No. 1 TS.
5. Correct a typographical error in Section 6.5.3.1.a of the
Millstone Unit No. 3 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 (SHC), which is provided below:
The proposed changes do not involve an SHC because the changes
do not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
No design basis accidents are affected by these proposed
changes. The proposed changes are administrative and editorial in
nature to reflect a recent reorganization, removal of the Executive
Vice President - Nuclear from receipt of PORC and SORC meeting
minutes, addition of the Senior Vice President - Millstone Station
to the receipt of PORC and SORC meeting minutes, and editorial
changes to the Millstone Unit Nos. 1 and 3 Technical Specifications.
No safety systems are adversely affected by the proposed changes,
and no failure modes are associated with the changes. Therefore,
there is no impact on the probability of occurrence or the
consequences of any design basis events.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
Since there are no changes in the way the plant is operated, the
potential for an unanalyzed accident is not created. There is no
impact on plant response, and no new failure modes are introduced.
These proposed administrative and editorial changes have no impact
on safety limits or design basis accidents, and they have no
potential to create a new or unanalyzed event.
3. Involve a significant reduction in a margin of safety.
The changes do not directly affect any protective boundaries nor
do they impact the safety limits for the protective boundaries.
These proposed changes are administrative and editorial in nature.
Therefore, there can be no reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, Connecticut 06360.
Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &
Howard, Counselors at Law, City Place, Hartford, Connecticut 06103-
3499.
NRC Project Director: John F. Stolz
Northeast Nuclear Energy Company, et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of amendment request: December 17, 1993
Description of amendment request: The proposed amendment would
revise the Technical Specifications to allow a relaxation in setpoint
tolerance of the pressurizer safety valves (PSVs) and main steam safety
valves (MSSVs) from plus or minus 1% to plus or minus 3% for the ``as-
found'' test condition.
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 (SHC), which is presented below:
The proposed changes do not involve an SHC because the changes
would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes revise the ``as found'' setpoint tolerances
for the PSVs and MSSVs from [plus or minus] 1% to [plus or minus]
3%. For the resetting of the PSVs and MSSVs, a [plus or minus] 1%
setpoint tolerance will be required prior to declaring the valve
operable for those instances where the [plus or minus] 1% tolerance
was exceeded. The proposed changes involve no hardware modifications
to plant structures, systems, or components. The proposed setpoint
tolerance of [plus or minus] 3% for the ``as-found'' condition was
previously evaluated as part of the PSE [Plant Safety Evaluation]
report for the transition to VANTAGE 5H fuel. The PSE was reviewed
and approved by the NRC staff as a part of a prior license
amendment.(9) In addition, since the proposed changes have
previously been evaluated by the PSE report, the calculated
radiological release associated with the PSE remain unaffected. In
addition, the proposed changes are in compliance with applicable
sections of the ASME Code and will not significantly affect
structural integrity of either the reactor coolant system or the
main steam system. Therefore, the proposed changes will have no
effect on the probability or consequences of previously evaluated
accidents.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes will not create the possibility of a new or
different kind of accident from those previously analyzed. The
changes revise the Technical Specifications so that setpoint
tolerance for the PSVs and MSSVs can be [plus or minus] 3% for the
``as-found'' condition. These changes have no effect on plant
operation. The PSV and MSSV setpoint drift in excess of the [plus or
minus] 1% lift setting is an occurrence which has previously and may
subsequently occur. The analyses for the transition to the VANTAGE
5H fuel have examined the effects on the plant accident analyses for
relaxation in PSV and MSSV setpoint tolerance to [plus or minus] 3%.
Also, these changes will have no effect on ASME Code compliance.
These changes do not introduce any new failures.
3. Involve a significant reduction in the margin of safety.
In support of the transition to the VANTAGE 5H fuel, a PSE was
performed which assumed a [plus or minus] 3% setpoint tolerance for
both the PSVs and MSSVs. Therefore, the effects of relaxing the PSV
and MSSV setpoints are already accounted for in the existing
analyses of record and will not affect the plants accident analyses.
Additionally, the proposed changes will have no significant effect
on the structural integrity of the reactor coolant system or the
main steam system. Also, for those occurrences where the ``as-
found'' setpoint of the PSV or MSSV is in excess of [plus or minus]
1%, a resetting to within [plus or minus] 1% of the valve setpoint
will be required prior to declaring the valve operable. Therefore,
the proposed changes will not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, Connecticut 06360.
Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &
Howard, City Place, Hartford, Connecticut 06103-3499.
NRC Project Director: John F. Stolz
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
DiabloCanyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis
Obispo County, California
Date of amendment requests: December 14, 1993 (Reference LAR 93-07)
T3Description of amendment requests: The proposed amendment would
revise the combined Technical Specifications (TS) for the Diablo Canyon
Power Plant (DCPP) Unit Nos. 1 and 2 to revise Technical Specification
(TS) 3/4.8.1, ``A.C. Sources'' to increase the required quantity of
emergency diesel generator (EDG) fuel oil stored in the engine-mounted
tank (day tank). The amendment request also proposes to revise TS 3/
4.7.11, ``Area Temperature Monitoring,'' and 3/4.8.1 to remove
references to a five EDG configuration. The specific TS changes
proposed are as follows:
(1) TS 3/4.7.11 would be revised to remove references to a common
(swing) diesel generator in Table 3.7-5.
(2) TS 3.8.1.1 and TS 3.8.1.2 would be revised to increase the
required minimum contained volume in the EDG engine-mounted fuel tank
(day tank) from 200 gallons to 250 gallons.
(3) TS 3.8.1.1 and TS 4.8.1.1.2 would be revised to remove
references to a five EDG configuration.
(4) TS 3.8.1.2 would be revised to correct a footnote. TS Bases 3/
4.8.1, 3/4.8.2, and 3/4.8.3 would be revised to clarify commitments to
Regulatory Guide 1.137 and expand the scope of information contained
within the TS Bases.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed increase in day tank TS minimum contained volume
requirements provides additional conservatism to assure the EDG fuel
oil contained in the day tank is sufficient to provide adequate time
for an operator to take corrective action to restore the fuel oil
supply to the affected day tank in the unlikely event that the fuel
oil supply from the main tanks were cut off.
Deletion of TS references to a five diesel generator
configuration and correction of the TS 3.8.1.2 footnote are
administrative changes that do not change the operating methodology
of DCPP. These proposed administrative changes remove outdated
information and correct an administrative oversight.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed increase in day tank TS minimum contained volume
requirements would not involve any physical change to the plant
systems or, in particular, to the EDG day tanks. The change does not
affect the ability of the EDGs to start and to fulfill their safety-
related function. Hence, no new failure mechanisms will be
introduced.
The proposed removal of references to a five EDG configuration
and correction of the TS 3.8.1.2 footnote are administrative in
nature. Further, the proposed changes would not result in any
physical alteration to any plant system. Therefore, the proposed
changes do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
Increasing the day tank TS minimum contained volume requirements
is a conservative change which provides additional margin to assure
the EDG fuel oil contained in the day tank is sufficient to provide
adequate time for an operator to take corrective action to restore
the fuel oil supply to the affected day tank in the unlikely event
that the fuel oil supply from the main tanks were cut off. The
proposed change will not alter any accident analysis assumptions,
initial conditions, or results. Consequently, the proposed change to
increase the EDG day tank TS contained fuel oil requirement does not
have any effect on the margin of safety.
The proposed administrative changes clarify the TS by removing
references to a five diesel generator configuration and correcting
the TS 3.8.1.2 footnote.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration. 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 NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120
NRC Project Director: Theodore R. Quay
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of amendment request: December 9, 1993
Description of amendment request: The amendment would change the
Operating Licenses and their corresponding Appendices A to reflect the
planned implementation of the Power Rerate Program at Limerick
Generating Station Units 1 and 2, and the corresponding increase in the
authorized maximum reactor core power level by five percent to 3458
megawatts thermal (MWt) from the current limit of 3293 MWt.
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 Operation License (OL) changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The proposed power rerate imposes only minor increases in the
plant operating conditions. Plant systems, components, and
structures have been verified to be capable of performing their
intended functions under rerated conditions. Where necessary, some
components will be modified or replaced prior to implementation of
the Power Rerate Program to accommodate the revised operating
condition. No new component or system interactions that could lead
to an accident are created. As discussed below, no transient events
result in a new sequence of events which could lead to a new
accident scenario. Emergency Core Cooling Systems (ECCS) - Loss-of-
Coolant Accident (LOCA) Analysis.
The current ECCS-LOCA performance analysis is already bounding
for power rerate conditions. The fuel peak cladding temperature for
rerate conditions is 1,345 deg.F, which is below the 2,200 deg.F
regulatory limit. Therefore, the analysis demonstrates that the LGS,
Units 1 and 2 will continue to comply with 10CFR50.46 and 10CFR50,
Appendix K.
Transient Event Analysis
The evaluation results for transient events indicate the margin
to the fuel Safety Limit Minimum Critical Power ratio (MCPR) will be
maintained for the 8x8 array fuel types, such as GE8x8NB or GE11
fuel design. The current fuel thermal-mechanical limits will
continue to be met.
Also, the power-dependent and flow-dependent MCPR and Maximum
Average Planar Linear Heat Generation Rate (MAPLHGR) limits
developed as part of the Average Power Range Monitor Rod Block
Monitor Technical Specifications (ARTS) improvement program are
applicable to power rerate. A TS Change Request to implement the
ARTS improvement program was submitted to the NRC by letter dated
August 27, 1993. The peak reactor vessel bottom head pressure will
remain within the American Society for Mechanical Engineers (ASME)
Code requirement for reactor overpressure protection.
The analysis performed focused on the most limiting transient
events in each disturbance category selected specifically for the
power rerate evaluations. The results demonstrated that LGS, Unit 1
and Unit 2 core thermal power output can be safely increased to
power rerate parameters without impacting plant safety during a
postulated transient event. The details of the impact to the
description in the UFSAR are delineated below.
a) Events Resulting in a Core Coolant Temperature Decrease
i) Loss of Feedwater Heating (LFWH)
The delta Critical Power Ratio (delta CPR) for the LFWH event at
the rerated power is bounded by the result estimated for the current
rated power level and remains significantly less than the Operating
Limit MCPR. There is no change between the delta CPR results for
high and low reactor core flow conditions. The calculated thermal
and mechanical overpowers for this event at power rerate conditions
also meet the fuel design criteria.
ii) Feedwater Controller Failure (FWCF) Maximum Demand
For the Increased Core Flow (ICF) and the Maximum Extended Load
Line Limit (MELLL) conditions, the trend for the FWCF - Maximum
Demand event at rerate conditions is consistent with the current
rated power analysis. For both high and low reactor core flow
conditions, the FWCF - Maximum Demand event becomes most limiting
due to the Turbine Bypass Valve Out-of-Service (TBVOOS) and the
Recirculation Pump Trip Out-of-Service (RPTOOS) analyses assumption.
The fuel thermal margin results remain within the acceptable limits
for the fuel type analyzed.
b) Events Resulting in a Reactor Pressure Increase
i) Turbine Trip with No Bypass (TTNBP)
At rerate conditions, the fuel transient thermal and mechanical
overpower results remain below the NRC acceptance criteria.
ii) Generator Load Rejection with No Bypass (LRNBP)
The fuel transient thermal responses are less severe than for
the TTNBP event described above. Therefore, at power rerate
conditions, the LRNBP event remains bounded by the TTNBP event.
iii) Main Steam Isolation Valve Closure, Flux Scram (MSIVF)
The peak reactor vessel bottom head pressure for rerate
conditions is slightly higher than the pressure at current rated
conditions due to the higher initial reactor coolant system
pressure. However, this result is still below the ASME overpressure
limit of 1,375 psig by a margin of 33 psi.
c) Events Resulting in a Core Coolant System Flow Rate Decrease
i) Recirculation Pump Seizure
The recirculation pump seizure assumes instantaneous stoppage of
the pump motor shaft of one recirculation pump. As a result, the
reactor core flow decreases rapidly. The reactor flow decreases
rapidly. The reactor vessel level swell due to the rapid reactor
core flow reduction reaches the high reactor water level setpoint,
causing a feedwater pump trip, a main turbine trip, and subsequently
a reactor scram on turbine stop valves closure. The peak neutron
flux and average fuel surface heat flux do not increase
significantly above the initial conditions, therefore no impact on
the fuel thermal margin is postulated to occur.
d) Events Resulting in Reactivity and Power Distribution
Anomalies
i) Rod Withdrawal Error (RWE)
The calculated delta CPR of 0.10 for this event at rerate
conditions is bounded by the generic ARTS - based RWE limits of
0.13. Therefore, the generic ARTS-based RWE analysis delta CPR
result is verified to be applicable for power rerate conditions for
LGS Units 1 and 2.
e) Events Resulting in a Reactor Coolant Inventory Increase
i) Inadvertent High Pressure Coolant Injection (HPCI) System
Actuation
Based on the peak average fuel surface heat flux results, the
HPCI actuation event will be bounded by the limiting pressurization
event (i.e., the TTNBP event described above) for delta CPR
consideration.
Anticipated Transients Without SCRAM (ATWS) Analysis
A generic evaluation for the ATWS event is provided in Section
3.7 of the Topical Report NEDC-31984P, ``Generic Evaluations of
General Electric Boiling Water Reactor Power Uprate,'' Supplement 1,
dated July 1991. This evaluation concludes that the ATWS acceptance
criteria for fuel, reactor pressure vessel (RPV) and containment
integrity will be met, if the following exists;
- Reactor power increases less than or equal to 5%
- Reactor Steam Dome pressure increases less than or equal to 40
psi;
- Safety Relief Valve (SRV) opening setpoints increase less than
or equal to 80 psi; and
- ATWS high pressure setpoint increases less than or equal to 20
psi.
The plant's parameter changes will remain within the above
criteria, except that the ATWS high pressure setpoint increase is 40
psi rather than 20 psi in order to maintain the same relationship
between the ATWS high pressure setpoint and the SRV opening
setpoints. Based on the previous analysis, this difference would
have a minor effect on the analysis results. The only significant
change is a slightly higher (i.e., about 10 psi) peak RPV pressure.
For additional assurance, a LGS specific ATWS analysis for a 5%
power rerate was performed. The events analyzed were:
1. Main Steam Isolation Valve (MSIV) Closure,
2. Pressure Regulator Failure - Open,
3. Loss of Feedwater, and
4. Inadvertent Opening of a Relief Valve.
The LGS specific analysis also concludes that the ATWS
acceptance criteria for fuel, RPV, and containment integrity will be
met for a 5% power rerate.
Other Evaluations
The impact of power rerate on the radiological consequences of
the accidents presented in UFSAR Chapter 15 was determined based on
the current design basis analyses, post rerate implementation system
conditions, and radiological source terms. In general, power rerate
will result in a small increase in the quantity of radioactive
material released during accidents and therefore slightly higher
(i.e., approximately 2% to 5%) accident doses. However, USFAR
Chapter 15 accident doses for rerated conditions remain within the
regulatory limits specified in 10CFR100 and 10CFR50, Appendix A, GDC
19.
The UFSAR Chapter 15 accidents that were evaluated and updated
for rerate conditions are as follows:
1) Loss of Coolant Accident (LOCA)
2) Main Steam Line Break (MSLB)
3) Fuel Handling Accident
4) Control Rod Drop Accident
5) Instrument Line Break
6) Feedwater Line Break
7) Steam Jet Air Ejector Line Break
8) Offgas System Failure
9) Liquid Radioactive Waste System Failure
An evaluation was also performed to address the power rerate
impact on accident mitigative features, structures, systems, and
components, within the balance of plant. The results are as follows:
- Auxiliary systems such as the Emergency Service Water,
Residual Heat Removal (RHR) Service Water, Ultimate Heat Sink (i.e.,
the spray pond), safety-related portions of secondary containment
reactor enclosure air cooling, primary containment drywell air
recirculation, and Emergency Diesel Generator enclosure ventilation
were confirmed to operate acceptably under normal and accident
conditions after implementation of power rerate.
- Combustible gas control systems were confirmed to be capable
of maintaining oxygen concentrations inside the primary containment
within regulatory limits under post accident rerate conditions.
- The secondary containment reactor enclosure recirculation
system and Standby Gas Treatment system were confirmed to be able to
adequately contain, process, and control the release of normal and
post-accident levels of radioactive material after implementation of
power rerate.
- Instrumentation was reviewed and confirmed to be capable of
performing their control and monitoring functions under rerate
conditions.
- Electric power systems including the main turbine generator
and switchgear components were verified as being capable of
providing the electrical load as a result of the rerated power
levels. No safety-related electrical loads were affected which would
impact the Emergency Diesel Generators.
- Piping systems were evaluated for the effect of operation at
higher power levels, including transient loadings. The evaluation
confirmed that with few exceptions piping and supports are adequate
to accommodate the increased loadings resulting from operation at
rerated power conditions. In a few cases, piping supports will be
modified to accept the higher forces due to rerate conditions.
- The effect of rerate conditions on high energy line break
(HELB) events for all Nuclear Steam Supply System (NSSS) and Balance
of Plant (BOP) systems was evaluated. The evaluation confirmed
structures, systems, and components important to safety are capable
of accommodating the effects of jet impingement and blowdown forces
and the environmental effects resulting from HELB events at rerate
conditions.
- The Moderate Energy Line Break (MELB) analysis was evaluated
for impact due to rerate conditions. Sufficient margin was
determined to exist in the original analysis to bound the rerate
conditions.
- Main control room (MCR) habitability was evaluated. Post-
accident MCR and Technical Support Center (TSC) doses were confirmed
to be within the limits of General Design Criterion (GDC) 19 of
10CFR50 Appendix A.
- Radiation doses for normal operation were reviewed and
confirmed to remain within the limits of 10CFR20 and 10CFR50,
Appendix I. The impact on post-accident sampling activities and
post-accident access to vital areas was also confirmed to be
acceptable.
- The environmental qualification of electrical and mechanical
equipment important to safety was evaluated for the impact of normal
and accident operating conditions at rerated power levels. The
majority of equipment will remain qualified for the new conditions.
For equipment that is not qualified, corrective actions will be
taken to ensure the plant equipment will perform their intended
functions under rerate conditions. No new equipment will be added
for power rerate which would increase the potential for component
failure. The Preventative Maintenance Program (PMP) will continue to
provide for appropriate equipment repair or replacement during
operation at rerated power conditions.
- The impact of operation at rerated power levels was evaluated
for Station Blackout and Fire Safety Shutdown area heat-up concerns.
The evaluation confirmed there is no adverse impact from rerate on
the ability of the plant to achieve safe shutdown under these
conditions.
- The consequences of postulated transients and special events
(i.e., ATWS and Station Blackout) will remain within NRC acceptance
criteria for rerate conditions. Concurrent malfunctions assumed to
occur during accidents have been accounted for in the safety
analyses for rerate conditions. The consequences of these equipment
malfunctions will not change with implementation of the Power Rerate
Program. Equipment that is important to safety either is capable of
or will be modified and/or replaced to be capable of performing its
intended function. The availability of redundant systems to provide
safety functions in the event of component malfunction is not
impacted as a result of rerate conditions. Furthermore, the impact
of power rerate on the consequences of abnormal transients and
accident conditions which are a result of component malfunctions has
been shown to be acceptable.
The probability (i.e., frequency of occurrence) of Design Basis
Accidents (DBAs) occurring is not affected by the proposed increased
power level, as the applicable regulatory criteria established for
plant equipment (e.g., ASME Code, the Institute of Electrical and
Electronics Engineers (IEEE) standards, National Electrical
Manufacturer's Association (NEMA) standards, NRC Regulatory Guides)
will still be followed as the plant is operated at the rerated power
level. Reactor SCRAM setpoints will be established such that there
is no significant increase in frequency due to rerate conditions. No
new challenges to safety-related equipment will result from the
implementation of power rerate.
The changes in consequences of hypothetical accidents which
would occur from 102% of the rerated power, compared to those
previously evaluated, are in all cases not significant, because the
accident evaluations from a power rerate to 105% of original rated
power will not result in exceeding the applicable NRC approved
acceptance limits. The spectrum of hypothetical accidents and
transients has been investigated, and has been determined to meet
the current regulatory criteria for LGS, Units 1 and 2 at rerate
conditions. The offsite radiological doses resulting from DBAs are
calculated to increase by only a few percent (i.e., approximately 2%
to 5%) because of the rerated power level, and will remain below
10CFR100 limits. In the area of reactor core design, the fuel
operating limits will continue to be met at the rerated power level,
and fuel reload analyses will continue to show that plant transients
will meet the criteria accepted by the NRC as specified in NEDO-
24011, ``GESTAR II.''
Challenges to fuel or ECCS performance were evaluated and shown
to still meet the criteria of 10CFR50.46 and 10CFR50, Appendix K.
Challenges to the primary containment have been evaluated and still
meet 10CFR50, Appendix A, GDC 38, ``Long Term Cooling,'' and GDC 50,
``Containment.'' Radiological release events have been evaluated and
have been shown to meet the guidelines of 10CFR100.
Therefore, the proposed OL changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2) The proposed OL changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
All actions to ensure that safety-related structures, systems,
and components will remain within their design allowable values, and
ensure that they can perform their intended functions under rerate
conditions will be taken prior to implementation of power rerate.
Power rerate does not increase challenges to or create any new
challenges to safety-related equipment or other equipment whose
failure could cause an accident. No new equipment is added as a
result of implementing the Power Rerate Program which would create
the possibility of a new type of accident. In addition, power rerate
does not create any new sequence of events or failure modes that
lead to a new type of accident.
Implementation of power rerate will increase the average neutron
flux in the reactor core, which increases the integrated neutron
fluence on the reactor pressure vessel (RPV) wall. To account for
the higher fluence, an RPV fracture toughness analysis was performed
for power rerate conditions. This analysis resulted in a proposed
revision to the ``pressure vs. temperature'' curves currently
provided in the Technical Specifications (TS), that will maintain
the current level of protection for the RPV. Therefore, power rerate
will not result in any new failure mode for the RPV, and thus, does
not create the possibility of a different type of accident from any
accident previously evaluated.
No new operating mode, safety-related equipment lineup, accident
scenario, or equipment failure mode was identified as resulting from
the implementation of the Power Rerate Program. The full spectrum of
accident considerations defined in NRC Regulatory Guide 1.70,
``Standard Format and Content of Safety Analysis Reports for Nuclear
Power Plants - LWR Edition,'' Revision 3, dated November 1978, have
been evaluated for rerate conditions and no new or different kind of
accident has been identified. Implementation of the Power Rerate
Program uses already-developed technology and applies it within the
capabilities of already existing plant equipment in accordance with
presently existing regulatory criteria to include applicable NRC
approved codes, standards, and methods. General Electric (GE) has
designed Boiling Water Reactors (BWRs) of higher power levels than
the rerated power of any of the currently operating BWR fleet and no
new power dependent accidents have been identified.
Therefore, the proposed OL changes do not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3) The proposed OL changes do not involve a significant
reduction in a margin of safety.
Power rerate will not involve a significant reduction in a
margin of safety, as plant equipment and reactions to transients and
hypothetical accidents will not result in exceeding the presently
approved NRC acceptance limits. The accident doses are calculated to
increase a few percent (approximately 2% to 5%) because of power
rerate, but remain below 10 CFR 100 limits. The events (i.e.,
transients, accidents, and ATWS) that form the bases of the TS were
evaluated for power rerate conditions. Although some changes to the
TS are required to implement power rerate, no NRC acceptance limit
will be exceeded. Therefore, the margins of safety with respect to
the safety limits and other TS bases will be maintained.
For systems addressed in the TS Section 2.2, 3/4.1, 3/4.2, 3/
4.3, 3/4.4, 3/4.5, 3/4.6 and 3/4.7 (i.e., Reactor Protection System,
Standby Liquid Control System, Power Distribution Limits,
Instrumentation, Reactor Coolant System, Emergency Core Cooling
Systems, Containment Systems, and Plant Systems), all components
will be operable and capable of performing their intended functions
under power rerate conditions such that the margin of safety is not
adversely impacted.
Therefore, the proposed OL changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101
NRC Project Director: Charles L. Miller
Philadelphia Electric Company, Public Service Electric and Gas
Company,Delmarva Power and Light Company, and Atlantic City
Electric Company,Dockets Nos. 50-277 and 50-278, Peach Bottom
Atomic Power Station,Units Nos. 2 and 3, York County, Pennsylvania
Date of application for amendments: December 21, 1993
Description of amendment request: The proposed Technical
Specification (TS) changes revise Table 3.2.F, ``Surveillance
Instrumentation,'' to accurately describe the main stack high range and
reactor building roof vent high range radiation monitors, and deletes
previously approved TS Change Request (TSCR) 91-10 for Unit 3 (License
Amendment No. 168). TSCR 91-10 requested an emergency temporary change
to the TS to allow fuel loading to take place without all control rods
fully inserted into the core.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Because the proposed changes are administrative in nature, they
do not affect the initial conditions or precursors assumed in the
Updated Final Safety Analysis Report Section 14. These changes do
not decrease the effectiveness of equipment relied upon to mitigate
the previously evaluated accidents.
Therefore, there is no increase in the probability or
consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed changes do not make any physical changes to the
plant or changes to operating procedures. Therefore, implementation
of the proposed changes will not affect the design function or
configuration of any component or introduce any new operating
scenarios or failure modes or accident initiation.
Therefore, the proposed changes do 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 administrative in nature and are
intended to provide clarification or eliminate confusion when
interpreting the Technical Specifications. The proposed changes do
not adversely affect the assumptions or sequence of events used in
any accident analysis.
Therefore, the proposed changes do not involve a 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,
Pennsylvania 17105
South Carolina Electric & Gas Company, South Carolina Public
ServiceAuthority, Docket No. 50-395, Virgil C. Summer Nuclear
Station, Unit No. 1, Fairfield County, South Carolina
Date of amendment request: October 29, 1993
Description of amendment request: The licensee is preparing to
replace the currently installed steam generators with new model Delta
75 steam generators (Delta 75 SGs). The new steam generators will be
larger than those currently installed. The physical changes to the
plant and the accident reanalyses needed to support those changes will
necessitate changes to the Technical Specifications (TS). The TS
changes requested involve alterations to the core operating limits,
changes to various reactor trip setpoints, deletion of the negative
flux rate trip, removal of references to specific analyses, changes to
the steam/feedwater flow mismatch activation setpoint, changes to
shutdown limits, changes to instrument uncertainty allowances, a change
to the methodology for reactor coolant system (RCS) flow determination,
modifications to departure from nucleate boiling (DNB) parameters, a
change to the engineered safety features actuation system setpoints for
steam generator water levels, removal of the F* and L* criteria, and
the addition of a requirement for a first inservice inspection for the
new steam generators. Due to the size of the new steam generators, TS
containing references to the maximum containment pressure following a
steam line break and the total RCS volume will also change; in
addition, a reference to RCS temperature is changed from a nominal
value to an indicated value.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), South Carolina Electric
& Gas Company (SCE&G or the licensee) has provided its analysis of the
issue of no significant hazards consideration, which is presented
below:
1) Operation of VCSNS [Virgil C. Summer Nuclear Station] in
accordance with the proposed license amendment does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
Implementation of the [Delta] 75 SGs and revised operating
conditions do not contribute to the initiation of any accident
evaluated in the FSAR [Final Safety Analysis Report]. Supporting
factors are as follows:
- The [Delta] 75 SG is designed in accordance with ASME
[American Society of Mechanical Engineers] Code Section III, 1986
edition [sic] and other applicable federal, state, and local laws,
codes and regulations and meets the original interfaces for the
Model D3 SGs with exception that provisions for a larger blowdown
nozzle have been made and the feedwater inlet nozzle is located in
the upper shell.
- All NSSS [nuclear steam supply system] components (i.e.,
reactor vessel, RC Pumps, pressurizer, CRDM's [control rod drive
mechanisms], [Delta] 75 SGs, and RCS piping) are compatible with the
revised operating conditions. Their structural integrity is
maintained during all proposed plant conditions through compliance
with the ASME code.
- Fluid and auxiliary systems which are important to safety are
not adversely impacted and will continue to perform their design
function.
- Overall plant performance and operation are not significantly
altered by the proposed changes.
Therefore, since the reactor coolant pressure boundary integrity
and system functions are not adversely impacted, the probability of
occurrence of an accident evaluated in the VCSNS FSAR will be no
greater than the original design basis of the plant.
An extensive analysis has been performed to evaluate the
consequences of the following accident types currently evaluated in
the VCSNS FSAR:
- Non-LOCA [loss-of-coolant accident]
- Large Break LOCA
- Steam Generator Tube Rupture
With the [Delta] 75 SGs and revised operating conditions, the
calculated results (i.e., DNBR [departure from nucleate boiling
ratio], Primary and Secondary System Pressure, Peak Clad
Temperature, Metal Water Reaction, Challenge to Long Term Cooling,
Environmental Conditions Inside and Outside Containment, etc.) for
the accidents are similar to those currently reported in the VCSNS
FSAR. Select results (i.e., Containment Pressure During a Steam Line
Break, Minimum DNBR for Rod Withdrawal from Subcritical, etc.) are
slightly more limiting than those reported in the current FSAR due
to the use of the assumed operating conditions with the new [Delta]
75 SGs, and in some cases, use of an uprated core power of 2900 MWt.
However, in all cases, the calculated results do not challenge the
integrity of the primary/secondary/ containment pressure boundary
and remain within the regulatory acceptance criteria applied to
VCSNS's current licensing basis. The assumptions utilized in the
radiological evaluations, described in Section 3.7, are thus
appropriate and are judged to provide a conservative estimate of the
radiological consequences during accident conditions. Given that
calculated radiological consequences are not significantly higher
than current FSAR results and remain well within 10CFR100 limits, it
is concluded that the consequences of an accident previously
evaluated in the FSAR are not increased.
2) The proposed license amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The [Delta] 75 SGs and revised operating conditions will not
introduce any new accident initiator mechanisms. Structural
integrity of the RCS is maintained during all plant conditions
through compliance with the ASME code. No new failure modes or
limiting single failures have been identified. Design requirements
of auxiliary systems are met with the RSGs [Replacement Steam
Generators]. Since the safety and design requirements continue to be
met and the integrity of the reactor coolant system pressure
boundary is not challenged, no new accident scenarios have been
created. Therefore, the types of accidents defined in the FSAR
continue to represent the credible spectrum of events to be analyzed
which determine safe plant operation.
3) The proposed license amendment does not involve a significant
reduction in a margin of safety.
Although the [Delta] 75 SGs and revised operating conditions
will require changes to the VCSNS Technical Specifications, it will
not invalidate the LOCA, non-LOCA, or SGTR [steam generator tube
rupture] conclusions presented in the FSAR accident analyses
(Appendix 6). For all the FSAR non-LOCA transients, the DNB design
basis, primary and secondary pressure limits, and dose limits
continue to be met. The LOCA peak cladding temperatures remain below
the limits specified in 10CFR50.46. The calculated doses resulting
from a SGTR event will continue to remain within a small fraction of
the 10CFR100 permissible releases. Environmental conditions
associated with High Energy Line Break (HELB) both inside and
outside containment have been evaluated. The containment design
pressure will not be violated as a result of the HELB. Equipment
qualification will be updated, as necessary, to reflect the revised
conditions resulting from HELB. The margin of safety with respect to
primary pressure boundary is provided, in part, by the safety
factors included in the ASME Code. Since the components remain in
compliance with the codes and standards in effect when VCSNS was
originally licensed (with the exception of the [Delta] 75 RSGs which
use the 1986 ASME Code Section III Edition), the margin of safety is
not reduced. Thus, there is no reduction in the margin to safety as
defined in the bases of the VCSNS Technical Specifications.
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: Fairfield County Library,
Garden and Washington Streets, Winnsboro, South Carolina 29180
Attorney for licensee: Randolph R. Mahan, South Carolina Electric &
Gas Company, Post Office Box 764, Columbia, South Carolina 29218
NRC Project Director: S. Singh Bajwa
South Carolina Electric & Gas Company, South Carolina Public
ServiceAuthority, Docket No. 50-395, Virgil C. Summer Nuclear
Station, Unit No. 1, Fairfield County, South Carolina
Date of amendment request: December 17, 1993
Description of amendment request: The proposed changes would revise
Technical Specification 3/4.3.3.6, ``Accident Monitoring
Instrumentation,'' and the associated Technical Specification Bases.
The changes are in accordance with the applicable guidance of Revision
3 to Regulatory Guide (RG) 1.97.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below. The proposed changes would
not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Regulatory Guide 1.97 furnishes standards acceptable to the NRC
for instrumentation to monitor plant variables and systems during
and following an accident. The purpose of the accident monitoring
instrumentation is to display plant variables that provide
information required by the control room operators for manual
actions and long term recovery. Determination of variable types and
category designations for VCSNS [Virgil C. Summer Nuclear Station]
was accomplished from a review of the Emergency Response Guidelines
(ERGs), the Final Safety Analysis Report, and the Westinghouse
Owners Group (WOG) ERGs. The WOG ERGs were used at VCSNS as a basis
for the Emergency Response Procedures. Operability of the
instruments used for accident monitoring ensures there is sufficient
information available on selected plant parameters to monitor plant
status during and following an accident. The changes proposed do not
effect components that can cause an accident. The increase in
allowable outage times from 7 to 30 days or from 48 hours to 7 days
does not significantly affect the consequences of an event
previously evaluated. The channel redundancy and the relatively
short outage times, coupled with the low probability of an event
requiring accident monitoring instrumentation during this interval,
ensure that sufficient information is available for operator manual
actions. The condition of the plant in either HOT STANDBY or HOT
SHUTDOWN, the first stage of the plant shutdown process, has no
impact on the assumptions made in the accident analysis.
The change in mode applicability for the Reactor Building Area
High Range Radiation Monitors to include modes 1, 2, and 3, but
exclude mode 4, is based on the usage of these monitors which is to
indicate a significant degradation of the reactor coolant pressure
boundary. These monitors do not initiate any automatic mitigation
system and are solely required to be operable to provide indication
which in conjunction with other operator actions will aid in
mitigating the consequences of design basis accidents. Design basis
accident sequences which may create a significant degradation of the
reactor coolant pressure boundary are not postulated to occur during
mode 4. Therefore, the proposed change does not increase the
probability or consequences of any accident previously evaluated.
(2) Create the possibility of a new or different kind of
accident from any previously evaluated.
The proposed change is consistent with the requirements of RG
1.97. The accident monitoring instrumentation will make available
reliable information to plant control room operators to mitigate the
consequences of a design basis accident. The first stages of plant
shutdown, HOT STANDBY and HOT SHUTDOWN, are plant modes for which
VCSNS has been analyzed. Since no plant configuration changes or
changes to the mode of operation of equipment, systems, and
components are introduced by the proposed Technical Specification,
no new failure modes or accident sequences are instituted.
Therefore, the changes proposed do not create the possibility of a
new or different kind of accident from any previously analyzed.
(3) Involve a significant reduction in a margin of safety.
The inclusion of category 1, type A or B, instrumentation in the
TS [Technical Specifications] provides assurance that adequate
information is available to the operators to maintain VCSNS in a
safe condition during and following a design basis accident.
Accomplishment of specific manual action by the control room
operators is enhanced due to the availability and reliability of the
indications. The proposed changes do not affect the design or
operation of safety related components relied upon to automatically
mitigate the consequences of a design basis event. The proposed
change from HOT SHUTDOWN to HOT STANDBY as the first stage of plant
shutdown will not affect the design or operation of any safety
related system or component. Therefore, the changes proposed would
not involve a 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: Fairfield County Library,
Garden and Washington Streets, Winnsboro, South Carolina 29180
Attorney for licensee: Randolph R. Mahan, South Carolina Electric &
Gas Company, Post Office Box 764, Columbia, South Carolina 29218
NRC Project Director: S. Singh Bajwa
Southern California Edison Company, et al., Docket Nos. 50-361 and
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3,
San Diego County, California
Date of amendment requests: November 3, 1993
T3Description of amendment requests: The licensee proposes to
revise the operability requirements of containment isolation valves
listed in Technical Specification (TS) Table 3.6-1, Section D. The
associated Bases 3/5.6.3, ``Containment Isolation Valves,'' is also
revised.
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. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: No
The proposed change provides new actions and Allowed Outage
Times (AOTs) for valves in Section D of Technical Specification (TS)
Table 3.6-1 that are currently allowed by the existing TS to be
secured for an indefinite period of time as long as they are secured
in their Engineered Safety Feature Actuation System (ESFAS) actuated
position. These valves are considered operable by the existing TS
although they may be unable to perform their containment isolation
function. The proposed change ensures that these valves are returned
to operable status within specified times based on the results of
specific risk evaluations on their contribution to core damage or
offsite dose release. The proposed change does not involve a
physical change to the facility as described in the Updated Final
Safety Analysis Report (UFSAR). Therefore, this proposed change does
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Will operation of the facility in accordance with this
proposed change create the possibility of a new or different kind of
accident from any accident previously evaluated?
Response: No
The ESFAS actuated positions of these valves are the positions
assumed in the safety analysis. There are no new accidents
associated with this proposed change because the previously analyzed
events already considered failures of containment isolation valves.
The plant is equipped with dual and redundant containment isolation
valves. Leaving the valves in their ESFAS actuated positions does
not create a new accident. Therefore, this proposed change does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
Response: No
This proposed change 1) limits the AOT of certain valves based
on contributions to core damage and offsite dose release when the
valves are secured in their ESFAS actuated position and 2) requires
these valves to be returned to OPERABLE status prior to Mode 4 entry
from a cold shutdown to ensure they are available to perform their
intended containment isolation function. Previously, these valves
could be secured in the ESFAS actuated position indefinitely.
Therefore, this proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Main Library, University of
California, P.O. Box 19557, Irvine, California 92713
Attorney for licensee: James A. Beoletto, Esquire, Southern
California Edison Company, P. O. Box 800, Rosemead, California 91770
NRC Project Director: Theodore R. Quay
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: December 27, 1993
Description of amendment request: The proposed change would revise
the Technical Specifications (TS) for the North Anna Power Station,
Units No. 1 and No. 2 (NA-1&2). The proposed changes revise the review
responsibilities of the Station Nuclear Safety and Operating Committee
(SNSOC) and the Management Safety Review Committee (MSRC).
The NA-1&2 TS address the organization and responsibilities of both
the onsite and offsite review groups: SNSOC and MSRC, respectively. The
responsibilities of the SNSOC include the review of new procedures and
changes to procedures that affect nuclear safety. The MSRC review
responsibilities include the review of safety evaluations and SNSOC
meeting minutes and reports. The extent of these review activities
would be revised by the proposed changes to ensure the two review
groups are focusing on nuclear safety issues and not spending an
unnecessary amount of time on activities of minimal safety
significance. Specifically, the proposed changes would revise the
review responsibilities of SNSOC regarding procedure changes. Rather
than reviewing all procedure changes, SNSOC would only review procedure
changes that require a safety evaluation. The proposed changes also
would revise the review responsibilities of the MSRC. Rather than
reviewing all of the safety evaluations and SNSOC meeting minutes and
reports as presently required by the TS, the MSRC would only review a
representative sample of these documents.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
[T]he elimination of the SNSOC review of procedure changes that
do not require a safety evaluation, revising the wording for
approval of procedure changes, and the modification of the MSRC's
duties regarding their review of safety evaluations and SNSOC
meeting minutes and reports will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated. As administrative
changes, the proposed Technical Specifications changes have no
direct or indirect effect on accident precursors. No plant
modifications are being implemented and operation of the plant is
unchanged. SNSOC review of new procedures and procedure changes that
require a safety evaluation ensures that activities that could
affect nuclear safety are being properly reviewed. The MSRC's
overview of representative samples of safety evaluations and SNSOC
meeting minutes and reports based on performance ensures these
programs are being properly implemented and nuclear safety is not
being compromised; or
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated since physical modifications
are not involved and systems and components will be operated as
before the change. The proposed changes are wholly administrative in
nature and have no impact on plant operations or accident
considerations. These changes modify the scope of SNSOC review of
procedure changes and MSRC's review functions concerning safety
evaluations and SNSOC meeting minutes and reports. Procedure changes
will continue to receive management review in accordance with
administative procedures, however, only changes that require a
safety evaluation will require SNSOC approval. MSRC review of
representative samples of safety evaluations and SNSOC meeting
minutes and reports based on performance will continue to provide
adequate assurance that nuclear safety is being properly considered;
or
3. Involve a significant reduction in a margin of safety as
defined in the basis of any Technical Specification since the
responsibilities of the SNSOC and MSRC are not addressed by the
existing Technical Specification Bases, nor are review requirements
for procedures. The proposed changes are administrative in nature
and have no impact on, nor were they considered in, existing UFSAR
accident analyses. Safety significant procedure changes, i.e.,
changes that require a safety evaluation to be prepared, will
continue to be reviewed by SNSOC, as will new procedures. Procedure
changes still require cognizant management approval and preparation
of an activity screening to determine whether or not the change
impacts nuclear safety. This ensures activities important to nuclear
safety are being appropriately reviewed. The effectiveness of the
safety evaluation program, and the thoroughness of SNSOC meetings
and reports will be assured through the MSRC's plant overview
function which is based on observed performance.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219
NRC Project Director: Herbert N. Berkow
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.
Duke Power Company, Docket No. 50-413, Catawba Nuclear Station,
Unit No. 1, York County, South Carolina
Date of amendment request: January 10, 1994
Description of amendment request: The proposed amendment would
revise Technical Specifications 2.0 and 3/4.2 which currently requires
the determination of the reactor coolant system flow rate by precision
heat balance measurement at least once per 18 months. Date of
publication of individual notice in Federal Register: January 26, 1994
(59 FR 3743)
Expiration date of individual notice: February 25, 1994
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Duke Power Company, Docket Nos. 50-269, 50-270 and 50-287, Oconee
Nuclear Station, Units 1, 2 and 3, Oconee County, South Carolina
Date of amendment request: November 11, 1993, as supplemented
November 22, 1993
Description of amendment request: The proposed amendments would
provide an interim acceptance criteria for control rod drop time on
Oconee Unit 1. Specifically, control rod Group 1, Rod 8, and Group 2,
Rod 5, would be considered operable with an insertion time of less than
or equal to 3.00 seconds provided that: (1) the average insertion time
for the remaining rods in Group 1 and the average insertion time for
the remaining rods in Group 2 is less than or equal to 1.5 seconds, and
(2) the core average negative reactivity insertion rate is within the
assumptions of the safety analysis. The acceptance criteria would apply
until the end of the current fuel cycle for Oconee Unit 1. This
acceptance criteria for rod drop time would apply for the two rods,
rather than the existing Technical Specification 4.7.1 limit of 2.00
seconds from the fully withdrawn position to 3/4 insertion.Date of
publication of individual notice in Federal Register: November 29, 1993
(58 FR 62689)
Expiration date of individual notice: December 29, 1993
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina 29691
Gulf States Utilities Company, Cajun Electric Power Cooperative,
and Entergy Operations, Inc., Docket No. 50-458, River Bend
Station, Unit 1, West Feliciana Parish, Louisiana
Date of application for amendment: December 8,
1993I11T3Brief description of amendment request: The
proposed amendment would grant one-time extensions for certain
Technical Specification surveillances which are currently required to
be performed beginning February 16, 1994. The licensee is requesting
extension of the surveillance intervals because the current operating
cycle has been extended, impacting the required completion dates for
these surveillances. Performance of these surveillances within the
required intervals would require that the plant be placed in an
undesirable operating configuration, or would necessitate a plant
shutdown. The surveillances for which extensions have been requested
will be performed during the fifth refueling outage, scheduled to begin
on April 16, 1994.
Date of individual notice in Federal Register: January 18, 1994 (59
FR 2630)
Expiration date of individual notice: February 17, 1994
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, Louisiana 70803
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: December 22, 1993
Description of amendment request: The proposed amendment would add
Limiting Conditions for Operation (LCO) and Surveillance Requirements
to Tables 3.12.1, ``Water Spray/Sprinkler Protected Areas,'' and
4.12.1, ``Water Spray/Sprinkler Tests,'' and clarify the associated
Bases to reflect the installation of a new full area fire suppression
system in the east and west cable tunnels. This new full area fire
suppression system was installed because the previous sprinkler system
did not provide coverage to some cable trays and the sprinkler head
orientation did not provide full coverage of the cable trays where it
was installed. The proposed amendment would also correct other portions
of Tables 3.12.1 and 4.12.1 for consistency with changes made to
reflect the east and west cable tunnel modification.Date of publication
of individual notice in Federal Register: January 18, 1994 (59 FR 2634)
Expiration date of individual notice: February 17, 1994
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.Notice Of Issuance Of Amendments To Facility Operating
Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC 20555, and at the local public document
rooms for the particular facilities involved.
Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: October 28, 1992, as
supplemented December 14, 1993
Brief description of amendments: The amendments remove Table 4.4-5,
``Reactor Vessel Material Surveillance Program Withdrawal Schedule,''
from the McGuire Technical Specifications and make other administrative
changes associated with the removal of the withdrawal schedule in
accordance with NRC Generic Letter 91-01.
Date of issuance: January 31, 1994
Effective date: January 31, 1994
Amendment Nos.: 139 and 121
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 23, 1992 (57
FR 61112) The December 14, 1993, letter provided clarifying information
that did not change the scope of the October 28, 1992, application and
the initial proposed no significant hazards consideration
determination.The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated January 31, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Astkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223
Duke Power Company, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South
CarolinaDate of application of amendments: July 14, 1993, as
supplemented August 24 and September 22, 1993
Brief description of amendments: The amendments revise TS 3.1.2.9
to clarify the role of High Pressure Injection and Core Flood Tank
deactivation in maintaining pilot operated relief valve operability for
low temperature overpressure protection (LTOP), add restrictions
regarding applicability of controls which assure 10 minutes are
available for operator action to mitigate an LTOP event, revise the
pressure-temperature limits and associated LTOP setpoints, and make
associated administrative changes. Also, the Bases would be revised to
be consistent with the above changes.
The conformance of the upper shelf energy and reactor vessel
material surveillance program to Appendices G and H will be determined
pending the NRC staff resolution of Generic Letter 92-01 in 1994.
Date of issuance: January 25, 1994
Effective date: To be issued within 30 days from the date of
issuance
Amendment Nos.: 204, 204, and 201
Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: September 1, 1993 (58
FR 46228) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated January 25, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina 29691
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412,
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,
Pennsylvania
Date of application for amendments: February 19, 1993
Brief description of amendments: The amendments revise the Appendix
A TSs 3.4.9.1, 3.4.9.2, and 4.4.9.2 relating to pressurizer surge line
stratification.
Date of issuance: January 31, 1994
Effective date: January 31, 1994
Amendment Nos.: 179 and 59
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: Aspril 28, 1993 (58 FR
25854) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated January 31, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear
One,Unit No. 2, Pope County, Arkansas
Date of application for amendment: February 24, 1993
Brief description of amendment: The amendment revised the
containment internal pressure lower limit of Technical Specification
Figure 3.6-1 from 12.8 to 13.2 psia.
Date of issuance: February 3, 1994
Effective date: 30-days from date of issuance
Amendment No.: 156
Facility Operating License No. NPF-6. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 31, 1993 (58 FR
16858) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 3, 1994. No significant
hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, Arkansas 72801.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251,
Turkey Point Plant Units 3 and 4, Dade County, Florida
Date of application for amendments: October 4, 1993
Brief description of amendments: The amendments revise the
surveillance test schedule in TS 4.6.1.2a and the associated Bases for
performing Type A test which determine the overall integrated
containment leakage rate.
Date of issuance: January 11, 1994
Effective date: January 11, 1994Amendment Nos. 158, 152Facility
Operating Licenses Nos. DPR-31 and DPR-41: Amendments revised the
Technical Specifications.
Date of initial notice in Federal Register: November 10, 1993 (58
FR 59748) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated January 11, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199.
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket No. 50-498, South Texas Project, Unit 1, Matagorda County,
Texas
Date of amendment request: December 6, 1993
Brief description of amendments: The amendments modify Technical
Specification 3.7.1.2 by extending the allowed outage time for the Unit
1 Train D turbine-driven auxiliary feedwater pump from 72 hours to 168
hours. This change is a one-time-only extension to accommodate an
augmented test program for the turbine driven auxiliary feedwater pump
during the restart of Unit 1 from the 1993 outage.
Date of issuance: January 25, 1994
Effective date: January 25, 1994, to be implemented within 10 days
of issuance.
Amendment No.: 58
Facility Operating License No. NPF-76. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 22, 1993 (58
FR 67848). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated January 25, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488
Illinois Power Company and Soyland Power Cooperative, Inc., Docket
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County,
Illinois
Date of application for amendment: November 4, 1993
Brief description of amendment: The amendment revises Clinton
Technical Specification 3/4.8.1.1, ``AC Sources - Operating,'' by
relocating the surveillance requirement to inspect the diesel
generators in accordance with the manufacturer's recommendations to the
preventive maintenance program.
Date of issuance: January 31, 1994
Effective date: January 31, 1994
Amendment No.: 87
Facility Operating License No. NPF-62. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 8, 1993 (58 FR
64610) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 31, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: The Vespasian Warner Public
Library District, 310 N. Quincy Street, Clinton, Illinois 61727.
Illinois Power Company and Soyland Power Cooperative, Inc., Docket
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County,
IllinoisDate of application for amendment: November 4, 1993
Brief description of amendment: The licensee proposed modifying
Technical Specification 3/4.8.2.1, ``DC Sources - Operating,'' by
deleting the requirement that the plant be shut down to perform the
required battery capacity or service testing. Following discussions
with the licensee, the staff has modified the licensee's proposal and
approved a one-time only change to permit replacement of the Division
IV battery subsystem at power.
Date of issuance: February 2, 1994
Effective date: February 2, 1994
Amendment No.: 88
Facility Operating License No. NPF-62. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 8, 1993 (58 FR
64610) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 2, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: The Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727.
Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook,
Nuclear Plant, Unit No. 2, Berrien County, Michigan
Date of application for amendment: Aspril 16, 1993, as supplemented
September 28 and December 3, 1993
Brief description of amendment: The amendment revises Technical
Specifications to allow certain tests normally designated as 18-month
surveillances to be delayed until the next refueling outage scheduled
to begin August 6, 1994. Extensions for four groups of surveillances
(Groups 1, 2, 6, 11) were previously approved for Unit 2 in Amendment
158 dated December 22, 1993. This amendment grants approval for the
extensions requested for the remaining 12 groups of surveillances and
completes the staff's review of the licensee's April 16, 1993 (as
supplemented) application.
Date of issuance: January 26, 1994
Effective date: January 26, 1994
Amendment No.: 159
Facility Operating License No. DPR-74. Amendment revises the
Technical Specifications. Dates of initial notice in Federal Register:
Asugust 4, 1993 (58 FR 41505) and December 21, 1993 (58 FR 67850)The
Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated January 26, 1994. No significant hazards
consideration comments received: No.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085.Northeast
Nuclear Energy Company, et al., Docket No. 50-336, Millstone Nuclear
Power Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: June 11, 1993, supplemented by
letter dated November 15, 1993
Brief description of amendment: The amendment revises the pressure/
temperature (P/T) limits for the reactor vessel. Specifically, Figure
3.4-2, ``Millstone Unit 2 Reactor Coolant System Presure-Temperature
Limitations for 12 Full Power Years,'' on page 3/4 4-19, is revised to
reflect the change in the curves and the title change to ``Millstone
Unit 2 Reactor Coolant System Pressure-Temperature Limitations for 20
EFPY.''
Date of issuance: January 27, 1994
Effective date: As of the date of issuance to be implemented
within30 days.
Amendment No.: 170
Facility Operating License No. DPR-65. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 21, 1993 (58 FR
39054) The November 15, 1993, submittal provided information that did
not change the initial proposed no significant hazards consideration
determination. The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 27, 1994. No significant
hazards consideration comments received: No.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, Connecticut 06360.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
DiabloCanyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis
Obispo County,California
Date of application for amendments: December 22, 1992, as
supplemented July 19, 1993 (Reference LAR 92-08)
Brief description of amendments: The amendments revise the combined
Technical Specifications (TS) for the Diablo Canyon Power Plant Unit
Nos. 1 and 2. Specifically, TS Section 3/4.3.2, ``Engineered Safety
Features Actuation System Instrumentation,'' would be revised to change
the second level undervoltage trip setpoint and allowable values.
Technical Specification 3/4.8.1, ``A.C. Sources,'' would also be
changed to revise the diesel generator (DG) steady state voltage
surveillance requirements. The second level undervoltage relay TS
setpoint and allowable values will be changed to maintain acceptable
voltages at the 480 volt and 120 volt buses during sustained degraded
voltage conditions. The DG steady state voltage surveillance
requirements will be changed to ensure that the diesel generators
provide adequate voltage when required to power the vital loads.
Date of issuance: January 6, 1994
Effective date: 60 days from date of issuance
Amendment Nos.: 86 & 85
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 3, 1993 (58 FR
7002) The July 19, 1993 submittal provided clarifying information and
did not affect the initial Federal Register notice and proposed no
significant hazards consideration. The Commission's related evaluation
of the amendments is contained in a Safety Evaluation dated January 6,
1994No significant hazards consideration comments received: No.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis
Obispo County,California
Date of application for amendments: July 6, 1993, as supplemented
December 29, 1993 (Reference LAR 93-03)
Brief description of amendments: The amendments revise the combined
Technical Specifications (TS) 3/4.3.2, ``Engineered Safety Features
Actuation System Instrumentation,'' Table 4.3-2, ``Engineered Safety
Features Actuation System Instrumentation Surveillance Requirements,''
for the Diablo Canyon Power Plant Unit Nos. 1 and 2 to relax the slave
relay test frequency for slave relays K612A, K614B, K615A, and K615B
from quarterly to once per 18 months during refueling or extended cold
shutdowns. The affected slave relays cause isolation of the charging
and letdown portions of the chemical and volume control system, and
actuate charging pump suction valves associated with volume control
tank and refueling water storage tank isolation.
Date of issuance: January 31, 1994
Effective date: For cycle 7 and after
Amendment Nos.: 87 and 86
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: Asugust 18, 1993 (58 FR
43929) The December 29, 1993, submittal provided clarifying information
and did not effect the initial Federal Register Notice and proposed no
significant hazards consideration.The Commission's related evaluation
of the amendments is contained in a Safety Evaluation dated January 31,
1994.No significan hazards consideration comments received: No.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407.
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: January 9, 1991, as
supplemented on August 19, 1991, June 22, 1992 and August 3, 1992
Brief description of amendments: The amendment changed the
Technical Specifications to revise the isolation setpoints for the
ambient temperature switches for the High Pressure Coolant Injection
and Reactor Core Isolation Cooling Systems room area coolers.
Date of issuance: January 31, 1993
Effective date: January 31, 1993
Amendment Nos.: 132 and 99
Facility Operating License Nos. NPF-14 and NPF-22. These amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 9, 1993 (58 FR
32389) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated January 31, 1993.No significant
hazards consideration comments received: No
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701.
Philadelphia Electric Company, Docket No. 50-353, Limerick
Generating Station, Unit 2, Montgomery County, Pennsylvania
Date of application for amendment: Asugust 27, 1993, as
supplemented November 10, and December 20, 1993
Brief description of amendment: The amendment allows a one-time TS
change to extend the allowed outage time (AOTs) for the Unit 2 residual
heat removal service water (RHRSW) system as well as the suppression
pool spray and suppression pool cooling modes of the residual heat
removal system from 72, 168 (i.e. seven days), and 72 hours,
respectively, to 288 hours (i.e., twelve days). The extended AOTs would
allow continued Unit 2 operation while maintenance isolation valves are
installed on both loops of the RHRSW system.
Date of issuance: January 26, 1994
Effective date: January 26, 1994
Amendment No. 30
Facility Operating License No. NPF-85. This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 29, 1993 (58
FR 50970) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 26, 1994. No significant
hazards consideration comments received: No
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Philadelphia Electric Company, Public Service Electric and Gas
CompanyDelmarva Power and Light Company, and Atlantic City Electric
Company,Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power
Station,Unit Nos. 2 and 3, York County, Pennsylvania
Date of application for amendments: October 5, 1993
Brief description of amendments: This amendment revised the Plant
Operating Review Committee review, the Nuclear Review Board review, the
Radiological Environmental Monitoring Program requirements, position
titles, and the organization chart in Appendix B of the Technical
Specifications (TS) to be consistent with Appendix A of the TS.
Date of issuance: January 26, 1994
Effective date: January 26, 1994Amendments Nos.: 183 and 188
Facility Operating License Nos. DPR-44 and DPR-56: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 8, 1993 (58 FR
64612) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated January 26, 1994. No significant
hazards consideration comments received: No
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of application for amendment: February 2, 1993, and
supplemented by letter dated November 16, 1993.
Brief description of amendment: The amendment extends the period of
time to reduce the setpoints of the Average Power Range Monitors and
the Rod Block Monitor when the plant enters single-loop operations.
Additionally, the change incorporates updated core values relative to
single loop operations and the addition of a new Specification 3.0.5
and its associated Bases.
Date of issuance: January 25, 1994
Effective date: As of date of issuance and shall be implemented
within 60 days of the date of issuance.
Amendment No.: 63
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 31, 1993 (58 FR
16872) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 25, 1994.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: May 21, 1993, as supplemented on
October 29, 1993, and November 16, 1993; the staff's proposed finding
of no significant hazards is not affected by these supplements.
Brief description of amendment: This amendment revises Technical
Specifications surveillance requirement 4.4.2.2 to apply only to the
pilot stage assembly of the safety relief valves (SRVs) and adds a new
surveillance requirement which will require the main portion of the
SRVs to be set pressure tested at least once every 5 years.
Date of issuance: January 27, 1994
Effective date: January 27, 1994
Amendment No.: 64
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: Asugust 18, 1993 (58 FR
43931) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 27, 1994.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: May 21, 1993 as supplemented on
August 23, 1993.
Brief description of amendment: This amendment revised a Technical
Specification surveillance requirement to increase the voltage limit
from 4580 to 4785 volts when performing the 18-month emergency diesel
generator full load rejection test.
Date of issuance: February 4, 1994
Effective date: Effective as of date of issuance and to be
implemented upon restart following fifth refueling outage currently
scheduled to begin on March 5, 1994.
Amendment No.: 65
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 23, 1993 (58 FR
34091) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 4, 1994. 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. 50-311, Salem
Nuclear Generating Station, Unit No. 2, Salem County, New Jersey
Date of application for amendment: Asugust 30, 1993
Brief description of amendment: The amendment changes the main
feedwater system containment isolation valves from the feedwater
control and control bypass valves to the feedwater stop check valves.
Date of issuance: January 21, 1994
Effective date: As of date of issuance and shall be implemented
within 60 days of the date of issuance
Amendment No. 128
Facility Operating License No. DPR-75: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 29, 1993 (58
FR50974) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 21, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, New Jersey 08079
Southern California Edison Company, et al., Docket Nos. 50-361 and
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3,
San Diego County, California
Date of application for amendments: November 25, 1991
Brief description of amendments: These amendments revise Technical
Specification (TS) 3/4.7.8, ``Fire Suppression Systems.'' This TS
revision deletes the phrase ``during shutdown'' from the fire pump
diesel engine surveillance requirement 4.7.8.1.2.c. This will allow the
surveillance of the fire pump diesel engine to be performed when one or
both Units 2 and 3 are in operation.
Date of issuance: February 1, 1994
Effective date: February 1, 1994
Amendment Nos.: 109 and 98
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 22, 1992 (57 FR
2600) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated February 1, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Main Library, University of
California, P. O. Box 19557, Irvine, California 92713.
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 application for amendments: September 30, 1993 (TS 345)
Brief description of amendment: The amendment deletes conditions
from the Browns Ferry Units 1, 2, and 3 licenses which require
maintenance of positive access controls for the containment in
accordance with 10 CFR 73.55(d)(8), and deletes a redundant condition
from the Unit 3 license.
Date of issuance: February 1, 1994
Effective date: February 1, 1994
Amendment Nos.: 202 - Unit 1; 221 - Unit 2; 175 - Unit 3
Facility Operating License Nos. DPR-33, DPR-52 and DPR-68:
Amendment revises the license conditions.
Date of initial notice in Federal Register: December 8, 1993 (58 FR
64616) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 1, 1994.No significant
hazards consideration comments received: None
Local Public Document Room location: Asthens Public Library, South
Street, Athens, Alabama 35611.
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit No. 1, Ottawa County, Ohio
Date of application for amendment: May 6, 1993
Brief description of amendment: The amendment revises the reporting
frequency requirements from semiannual to annual for submission to the
NRC of the Radioactive Effluent Release Report, and clarifies the
reporting requirements regarding steam generator tube inspection
Category C-3 results.
Date of issuance: December 30, 1993
Effective date: December 30, 1993
Amendment No.: 184
Facility Operating License No. NPF-3. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 23, 1993 (58 FR
34096) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 30, 1993.No significant
hazards consideration comments received: No
Local Public Document Room location: University of Toledo Library,
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia.
Date of application for amendments: July 16, 1993, as supplemented
November 15, 1993. The November 15, 1993, submittal did not expand the
scope of the original application and did not change the proposed no
significant hazards consideration determination.
Brief description of amendments: These amendments implement the
revised 10 CFR Part 20, Standards for Protection Against Radiation, and
reflect revisions to 10 CFR 50.36a.
Date of issuance: January 25, 1994
Effective date: January 25, 1994
Amendment Nos. 185 and 185
Facility Operating License Nos. DPR-32 and DPR-37: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: Asugust 18, 1993 (58 FR
43937) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 25, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia.
Date of application for amendments: March 19, 1993, as supplemented
December 9, 1993.
Brief description of amendments: These amendments address plant
operation with a control rod urgent alarm failure, a change in the
control rod assembly partial movement surveillance test frequency, and
proposed administrative changes.
Date of issuance: February 4, 1994
Effective date: February 4, 1994
Amendment Nos. 186 and 186
Facility Operating License Nos. DPR-32 and DPR-37: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 12, 1993 (58 FR
28064) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 4, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301
Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks,
Manitowoc County, Wisconsin
Date of application for amendments: January 14, 1993
Brief description of amendments: The amendments split Technical
Specification (TS) 15.3.1.E.2, which defines the allowable limits of
chloride and fluoride in the reactor coolant, into two individual
Limiting Conditions for Operation (LCOs), thus clarifying the reactor
coolant chemistry limitations. In addition, the amendments added a 24-
hour hot shutdown action statement to the reactor coolant impurity
limit LCOs. The amendments also modified the corresponding TS Bases
Section.
Date of issuance: January 27, 1994
Effective date: January 27, 1994
Amendment Nos.: 145 and 149
Facility Operating License Nos. DPR-24 and DPR-27: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 3, 1993 (58 FR
12270) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated January 27, 1994. No significant
hazards consideration comments received: No.
Local Public Document Room location: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241.
Notice Of Issuance Of Amendments To Facility Operating Licenses And
Final Determination Of No Significant Hazards Consideration And
Opportunity For A Hearing (Exigent Public Announcement Or Emergency
Circumstances)
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 for the
amendment 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.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective 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
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC 20555,
and at the local public document room for the particular facility
involved.
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By March 18, 1994, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.714 which is available at the
Commission's Public Document Room, the Gelman Building, 2120 L Street,
NW., Washington, DC 20555 and at the local public document room for the
particular facility involved. If a request for a hearing or petition
for leave to intervene is filed by the above date, the Commission or an
Atomic Safety and Licensing Board, designated by the Commission or by
the Chairman of the Atomic Safety and Licensing Board Panel, will rule
on the request and/or petition; and the Secretary or the designated
Atomic Safety and Licensing Board will issue a notice of a hearing or
an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination
that the amendment involves no significant hazards consideration, if a
hearing is requested, it will not stay the effectiveness of the
amendment. Any hearing held would take place while the amendment is in
effect.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington, DC 20555, by the
above date. Where petitions are filed during the last 10 days of the
notice period, it is requested that the petitioner promptly so inform
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and
to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
Gulf States Utilities Company, Cajun Electric Power Cooperative,
and Entergy Operations, Inc., Docket No. 50-458, River Bend
Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: November 18, 1993, as supplemented by
letter dated December 21, 1993.
Brief description of amendment: The amendment revises the River
Bend, Unit 1 Technical Specifications to permit extending the time to
perform leak rate testing of certain containment isolation valves and
pressure isolation valves so that the testing can be performed during
the refueling outage scheduled to start April 16, 1994, rather than
requiring an earlier shutdown solely to perform the testing. Also, an
exemption to 10 CFR Appendix J was issued on February 2, 1994, that
provides an extension, consistent with the revision to the technical
specifications, to allow the testing of containment isolation valves to
be delayed until the refueling outage.
Date of issuance: February 2, 1994
Effective date: February 2, 1994
Amendment No.: Asmendment No. 71
Facility Operating License No. NPF-47: The amendment revised the
Technical Specifications. Public comments requested as to proposed no
significant hazards consideration: Yes. January 5, 1994 (59 FR 616)The
Commission's related evaluation of the amendment, finding of emergency
circumstances, and final determination of no significant hazards
consideration are contained in a Safety Evaluation dated February 2,
1994.
Attorney for licensee: Mark Wetterhahn, Esq., Bishop, Cook, Purcell
and Reynolds, 1401 L Street, NW., Washington, D.C. 20005
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, Louisiana 70803.
Pennsylvania Power and Light Company, Docket No. 50-388,
Susquehanna Steam Electric Station, Unit 2, Luzerne County,
Pennsylvania
Date of application for amendment: January 24, 1994
Brief description of amendment: The amendment revised the
applicability requirement in Sections 3.0.4, 4.0.4, 3.3.7.5 Action 80,
4.3.7.5, 3.4.2 Action c, and 4.4.2 of the Technical Specifications to
permit Susquehanna, Unit 2 to continue to operate with the acoustic
monitor on the ``S'' safety/relief valve tailpipe inoperable.
Date of issuance: January 31, 1994
Effective date: As of its date of issuance and will remain in
effect until the next shutdown of sufficient duration to allow for
containment entry, not to exceed the sixth refueling and inspection
outage.
Amendment No.: 100
Facility Operating License No. NPF-22: Amendment revised the
Technical Specifications. Public comments requested as to proposed no
significant hazards consideration: No. On January 27, 1994, the staff
issued a Notice of Enforcement Discretion, which was immediately
effective and remained in effect until this amendment was issued.
The Commission's related evaluation of the amendment, finding of
emergency circumstances, consultation with the Commonwealth of
Pennsylvania and final no significant hazards considerations
determination are contained in a Safety Evaluation dated January 31,
1994.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts &
Trowbridge 2300 N Street NW., Washington, D.C. 20037
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18071.
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of application for amendment: January 13, 1994
Brief description of amendment: The amendment modified the
Technical Specifications (TS) to defer response time testing for low
pressure emergency core cooling systems (ECCS) until startup following
the next cold shutdown, but not later than the startup following
completion of the spring 1994 refueling outage.
Date of issuance: January 31, 1994
Effective date: January 31, 1994
Amendment No.: 120
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications. Public comments on proposed no significant
hazards consideration comments received: No. The Commission's related
evaluation of the amendment, finding of emergency circumstances, and
final determination of no significant hazards consideration are
contained in a Safety Evaluation dated January 31, 1994.
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352
Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005-3502.
NRC Project Director: The odore R. Quay
Dated at Rockville, Maryland, this
For the Nuclear Regulatory Commission
Robert A. Capra,
Acting Director, Division of Reactor Projects -I/II, Office of Nuclear
Reactor Regulation
[Doc. 94-3465 Filed 2-15-94; 8:45 am]
BILLING CODE 7590-01-F
