# Biweekly Notice

> Briefs, arguments, decisions, and more.

URL: https://www.frixlaw.com/law-library/documents/fr%3AX94-11012

## Record

- **Collection:** Federal Register
- **Document type:** Uncategorized Document
- **Published:** October 12, 1994

## Text

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating Licenses
Involving No Significant Hazards Considerations

I. Background

Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from September 19, 1994, through September 29,
1994. The last biweekly notice was published on September 28, 1994 (59
FR 49425).

Notice Of Consideration Of Issuance Of Amendments To Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, And Opportunity For A Hearing

The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, and should cite the publication date and page
number of this Federal Register notice. Written comments may also be
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the NRC Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC
20555. The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By November 14, 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.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,
North Carolina

Date of amendment request: August 19, 1994
Description of amendment request: The proposed amendment will move
the current procedural details of the radiological effluent Technical
Specifications (TS) programmatic controls for radioactive effluents,
radiological environmental monitoring and solid radioactive wastes from
the Administrative Controls Section of the TS to the Offsite Dose
Calculation Manual (ODCM) or the Process Control Program (PCP), as
appropriate, in accordance with the guidance of Generic Letter 89-01.
This amendment will also incorporate changes to the reporting
requirements for the Effluent Release Reports, in accordance with 10
CFR 50.36; incorporate references to the new 10 CFR Part 20; and revise
the terminology for the gaseous effluent release rate limits.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Transferring the procedural details from the TS to the ODCM and
PCP and their replacement with programmatic controls have no impact
on plant operation or safety. No safety-related equipment, safety
function, or plant operation will be altered as a result of this
proposed change. The changes are unrelated to the initiation and
mitigation of accidents and equipment malfunctions addressed in the
Final Safety Analysis Report.
The proposed revisions to the reporting requirements for
Effluent Release Reports, the gaseous effluent release rate limit
and the relocation of the old 10 CFR 20.106 requirements to the new
10 CFR 20.1302 have no impact on plant systems, plant operations or
accident precursors. The changes to the Effluent Report requirements
and the updated reference to 10 CFR 20.1302 are administrative in
nature. The change to the gaseous effluent release limit is also
administrative in nature in that it will allow the continued
operation of the facility with the same release rate limits as are
currently implemented by the Technical Specifications.
Therefore, there would be no increase in the probability or
consequences of an accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Transferring the procedural details from the TS to the ODCM and
PCP and their replacement with programmatic controls have no impact
on plant operation or safety. No safety-related equipment, safety
function, or plant operation will be altered as a result of this
proposed change. No changes to plant components or structures are
introduced which could create new accidents or malfunctions not
previously evaluated.
The proposed revisions to the reporting requirements for
effluent Release Reports, the gaseous effluent release rate limit
and the relocation of the old 10 CFR 20.106 requirements to the new
10 CFR 20.1302 have no impact on plant systems, plant operations or
accident precursors. The changes to the Effluent Report requirements
and the updated reference to 10 CFR 20.1302 are administrative in
nature. The change to the gaseous effluent release limits is also
administrative in nature in that it will allow the continued
operation of the facility with the same release rate limits as are
currently implemented by the Technical Specifications.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident
previously evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
The procedural details of the current RETS will be transferred
to the ODCM and PCP and replaced with programmatic controls
consistent with regulatory requirements, including controls on
revisions to the ODCM and PCP. Thus, no requirements or controls
will be reduced.
The changes to the Effluent Report requirements and the updated
reference to 10 CFR 20.1302 are administrative in nature and
therefore have no effect on the margin of safety. The proposed
revisions to the gaseous effluent release limits will maintain the
release rate limits at the same level as currently implemented by
the Technical Specifications. Therefore, there will be no change in
the types and amounts of effluents that will be released, nor will
there be an increase in individual or cumulative radiation exposures
to any member of the public.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
NRC Project Director: David B. Matthews

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station, Units 1 and 2, Lake County, Illinois

Date of amendment request: September 19, 1994
Description of amendment request: The proposed amendment would
revise the Technical Specifications by reducing the frequency for
testing the containment spray system spray nozzles.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability of occurrence or consequences of any accident
previously evaluated.
The relaxation of surveillance frequency will not affect any of
the initiators or precursors of any accident previously evaluated.
Performance of CS spray nozzle testing on a ten year basis rather
than on a five year basis will not increase the likelihood that a
transient initiating event will occur because transients are
initiated by external events, equipment malfunction, and/or
catastrophic system failure. There are no failure mechanisms or
modes for the CS system or spray nozzles that could initiate a
transient since the CS system is passive except during a Loss of
Coolant Accident (LOCA). Upon receipt of a Containment Spray signal
(Containment High-High Pressure coincident with a Safety Injection
Signal), the CS pumps automatically start and valves align to
provide spray flow through the CS risers, ring headers, and out the
spray nozzles. Periodic testing requirements for the CS pumps and
valves (the active components of the system) are unaffected by the
proposed changes. Industry experience and previous test experience
at Zion Station supports the conclusion that functional checks of
the spray nozzles on a ten year basis is adequate to detect
degradation or blockage of the spray nozzles.
The proposed typographical and administrative changes do not
affect the operability or surveillance requirements given in
Technical Specifications. They will only improve consistency of
existing terminology and format of Technical Specifications and will
remove temporarily imposed Bases that are no longer applicable.
Based on the fact that reliability of the system will not be
affected and transient precursors and initiators are not affected by
operation in accordance with the proposed changes, the probability
of occurrence of accidents previously evaluated will not
significantly increase.
The proposed change in surveillance frequency will not affect
the ability of the CS system to function as designed during the
accidents considered in the Safety Analyses. Periodic testing
requirements for the CS pumps and valves (the active components of
the system) are unaffected by the proposed changes. Industry
experience and previous test experience at Zion Station supports the
conclusion that functional checks of the spray nozzles on a ten year
basis is adequate to detect degradation or blockage of the spray
nozzles. Given the proposed changes, the CS system will maintain the
ability to reduce containment pressure, remove heat from
containment, and remove iodine from the containment atmosphere
during the design basis LOCA. As a result, peak containment pressure
will be maintained below design pressure and the off-site release
due to the postulated accident will remain as described in the
Safety Analyses. Therefore, based on the previous discussion, the
proposed changes do not involve a significant increase in
consequences of any accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any previously analyzed.
The proposed changes to the Technical Specifications do not
involve the addition of any new or different types of safety related
equipment, nor does it involve the operation of equipment required
for safety operation of the facility in a manner different from
those addressed in the safety analyses. No safety related equipment
or function will be altered as a result of the proposed changes.
Also, changes to the procedures governing normal plant operation and
recovery from an accident are not necessitated by the proposed
Technical Specification changes.
The proposed typographical and administrative changes do not
affect the operability or surveillance requirements given in
Technical Specifications. They will only improve consistency of
existing terminology and format of Technical Specifications and will
remove Bases that are no longer applicable.
Since no new failure modes or mechanisms are added by the
proposed changes, the possibility or a new or different kind of
accident is not created.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
Plant safety margins are established through LCOs, limiting
safety system settings, and safety limits specified in the Technical
Specifications. There will be no changes to either the physical
design of the plant or to any of these settings and limits as a
result of relaxing the surveillance frequency of CS nozzle checks
from five years to ten years. Testing on a ten year basis is
adequate to detect spray nozzle degradation or blockage since the
system piping and spray nozzles are constructed of corrosion
resistant Type 304 stainless steel and since the system is normally
passive (i.e. spray risers and spray rings are empty with no flow
except during an accident). This conclusion was also provided in
NUREG-1366 and Generic Letter 93-05 which recommended revising the
surveillance frequency as proposed.
The proposed typographical and administrative changes do not
affect the operability or surveillance requirements given in
Technical Specifications. They will only improve consistency of
existing terminology and format of Technical Specifications and will
remove Bases that are no longer applicable.
Based on the above discussion, the ability to safely shut down
the operating unit and mitigate the consequences of all accidents
previously evaluated will be maintained. Therefore, the margin of
safety is not significantly affected.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Waukegan Public Library, 128
N. County Street, Waukegan, Illinois 60085
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690
NRC Project Director: Robert A. Capra

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: August 25, 1994
Description of amendment request: The requested amendments modify
the trip setpoint and allowable value for the 4 kilo-volt (KV)
electrical bus degraded grid undervoltage relay and the allowable value
for the loss of offsite power relay in response to an issue identified
in the licensee's Self-Initiated Technical Audit.
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. The amendments will not affect either the probability or
the consequences of an accident, since no physical changes to the
plant are being proposed. The amendments merely change the existing
technical specification settings for the above relays to more
conservative values. Current field settings for these relays are
already at these more conservative values. No changes to the manner
in which the plant is operated are being proposed.
Criterion 2
The requested amendments will not create the possibility of a
new or different kind of accident from any accident previously
evaluated. As stated above, no actual changes to the physical plant
are being proposed. No effect on plant operation will occur,
therefore the possibility of new accident types is not created.
Criterion 3
The requested amendments will not involve a significant
reduction in a margin of safety. Plant safety margins will be
unaffected, since no changes to the plant are being made. The
proposed technical specification values are more conservative and
are intended to make the technical specifications correspond with
the actual plant relay settings.
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: Herbert N. Berkow

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: August 25, 1994
Description of amendment request: The amendments would change the
frequency for conducting the surveillance test required by TS 4.7.1.2.1
for the auxiliary feedwater pumps from once per 31 days to at least
once per 92 days and would add a footnote which clarifies that testing
is not required to be performed until system heatup has progressed to a
pressure (600 psig) that will support conduct of the test. The change
in the surveillance frequency has been evaluated and approved by the
NRC staff as discussed in Section 9.1 of NUREG-1366, ``Improvements to
Technical Specifications Surveillance Requirements.'' The change is
based on the finding in NUREG-1366 that an analysis of AFW pump
failures indicates that a monthly surveillance test interval may be
contributing to AFW pump unavailability through failures and equipment
degradation and, therefore, AFW pump availability is increased by
quarterly testing on a staggered basis. Generic Letter 93-05, ``Line-
Item Technical Specification Improvements to Reduce Surveillance
Requirements for Testing During Power Operation,'' provided the sample
TS for this change. The change is accomplished by dividing TS
4.7.1.2.1a into two parts. The new 4.7.1.2.1a maintains the previous
31-day testing frequency for the AFW valves while the new 4.7.1.2.1b
inserts a new frequency of once per 92 days for the AFW pump tests.
Also, an obsolete footnote is deleted. The new footnote discussed above
is consistent with NUREG-1431, ``Standard Technical Specifications for
Westinghouse Plants.'' Appropriate changes to the Bases for the TS have
also been proposed.
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 requested amendments decrease from monthly to quarterly the
frequency at which the motor-driven and turbine-driven AFW pumps
must be demonstrated operable as specified in TS 4.7.1.2.1. They
also incorporate a note of clarification from the new Westinghouse
STS into the existing Catawba specifications concerning when the
pump head or discharge pressure versus flow verification for the
turbine-driven pump is required to be performed.
Criterion 1
The requested amendments will not involve a significant increase
in the probability or consequences of an accident previously
evaluated. Decreasing the frequency of AFW pump testing as specified
in TS from monthly to quarterly will have no impact upon the
probability of any accident, since the AFW pumps are not accident
initiating equipment. Also, since Catawba's AFW pump performance
history supports making the proposed change, system response
following an accident will not be adversely affected. Therefore, the
requested amendments will not result in increased accident
consequences. Deletion of the obsolete footnotes as indicated in the
Catawba technical specification markups is purely an administrative
change, and therefore will have no impact upon either the
probability or consequences of any accident. Incorporating the new
STS note will only serve to clarify when the turbine-driven pump is
required to be tested and will not have any impact upon either the
probability or consequences of any accident. The pump will still be
tested as before and its acceptance criteria will be unaffected.
Criterion 2
The requested amendments will not create the possibility of a
new or different kind of accident from any accident previously
evaluated. As stated above, the AFW pumps are not accident
initiating equipment. No new failure modes can be created from an
accident standpoint. The plant will not be operated in a different
manner. Deletion of the Catawba obsolete footnotes has no bearing on
any accident initiating mechanisms. Incorporating the clarifying
note from the new STS will not result in any new acident sequences,
since plant operation will be unaffected.
Criterion 3
The requested amendments will not involve a significant
reduction in a margin of safety. Plant safety margins will be
unaffected by the proposed changes. The AFW pumps will still be
capable of fulfilling their required safety function, since plant
operating experience supports the proposed change. The availability
of the AFW pumps will be increased as a result of the proposed
amendments because they will not have to be made unavailable for
testing as frequently. Finally, the proposed amendments are
consistent with the NRC position and guidance set forth in NUREG-
1366 and Generic Letter 93-05. Deletion of the Catawba obsolete
footnotes will not result in any impact to plant safety margins.
Incorporating the note from the new STS will not impact any safety
margins.
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: Herbert N. Berkow

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile
Point Nuclear Station Unit No. 1, Oswego County, New York

Date of amendment request: September 1, 1994
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.2.2, ``Minimum Reactor Vessel
Temperature for Pressurization,'' and the associated Bases.
Specifically, the proposed amendment replaces existing TS Figures
3.2.2.a,b,c,d, and e and associated TS Tables 3.2.2.a,b,c,d, and e,
that define the limits for minimum reactor vessel temperature for
pressurization and account for neutron damage at exposures up to 18
effective full power years (EFPY), with new figures and tables that are
applicable for up to 18 EFPY. The licensee stated that the new
pressure-temperature (P-T) limits were developed based on a plant-
specific Charpy shift model for Nine Mile Point Nuclear Station Unit
No. 1 which is consistent with and meets the requirements of Regulatory
Guide 1.99, Revision 2, ``Radiation Embrittlement of Reactor Vessel
Materials.'' The new P-T limits were calculated in accordance with 10
CFR Part 50, Appendix G, and with the requirements specified in
Appendix G to Section III of the American Society of Mechnical
Engineers Boiler and Pressure Vessel Code (ASME Code).
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 [NMP1], in accordance
with the proposed amendment, will not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Components of the reactor primary coolant system are operated so
that no substantial mechanical or thermal loading is applied unless
the reactor pressure vessel (RPV) materials are at a temperature
well above the reference nil-ductility temperature (RTNDT) of
the limiting RPV material. Protection against brittle fracture is
further ensured by postulating a defect with a depth 1/4 of the RPV
wall thickness and a length 1-1/2 times the wall thickness, and
calculating the allowable pressure loading as a function of
temperature using linear elastic fracture mechanics. Safety factors
are applied to the allowable loading determination and lower bound
fracture toughness properties are used to represent the material
behavior. The net effect of the 10 CFR [Part] 50, Appendix G and the
ASME Section III, Appendix G P-T curve calculative procedures is to
produce very conservative P-T curves. These procedures have been
applied in the calculation of the proposed P-T limits.
Neutron damage during plant operation is accounted for in the
allowable pressure loading by calculating an adjusted reference nil-
ductility temperature (ARTNDT). Regulatory Guide 1.99, Revision
2, defines the ARTNDT as the sum of the reference nil-ductility
temperature (RTNDT) plus the shift in the reference nil-
ductility temperature caused by irradiation ([delta]RTNDT),
plus a margin. The proposed amendment replaces Equation (2) in
Regulatory Position 2.1 with an accurate plant-specific model. The
ARTNDT margin is the same as for earlier P-T curve
calculations. Operation of NMP1 in accordance with the proposed P-T
operating limits will preclude brittle failure of the RPV materials.
Safety margins for brittle fracture are in accordance with those
specified in 10 CFR [Part] 50, Appendix G and Appendix G to Section
III of the ASME Code. Therefore, 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 accident previously evaluated.
The proposed amendment incorporates P-T operating limits based
on previously established calculative procedures described in 10 CFR
[Part] 50, Appendix G, Appendix G to Section III of the ASME Code,
and Regulatory Guide 1.99, Revision 2. The proposed changes to the
P-T operating limits are based on analyses of the irradiated
limiting plate material for Nine Mile Point Unit 1. The proposed
changes do not modify any plant equipment nor do they create any
potential initiating events that would create any new or different
kind of accident. Operation in accordance with the proposed P-T
operating limits will preclude brittle failure of the reactor vessel
material, since safety margins specified in 10 CFR [Part] 50,
Appendix G and Appendix G to Section III of the ASME Code will be
maintained. Therefore, the proposed P-T limits will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The operation of Nine Mile Point Unit 1, in accordance with the
amendment, will not involve a significant reduction in a margin of
safety.
Operation in accordance with the proposed P-T operating limits
will preclude brittle failure of the reactor pressure vessel since
safety margins in 10 CFR [Part] 50, Appendix G and Appendix G to
Section III of the ASME Code will be maintained. The plant-specific
limiting material [delta]RTNDT has been reduced as compared
with the overly conservative [delta]RTNDT used in previous P-T
curve calculations as a result of the more accurate representation
of the Nine Mile Point Unit 1 RPV plate behavior as a function of
neutron exposure. However, the [delta]RTNDT is intended to be
an accurate representation of the Charpy shift (indexed at 30 ft-lbs
of absorbed energy) as a function of fluence. Since the ASME Section
III, Appendix G safety factors have been maintained and the
Regulatory Guide 1.99, Revision 2, margin term specified in
Regulatory Position 2.1 has been applied in the same manner as in
earlier P-T curve calculations, no significant reduction in the
margin of safety has resulted from the use of a plant-specific
[delta]RTNDT model. Therefore, the proposed amendment will not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: Michael J. Case

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 amendment request: August 17, 1994 (Reference LAR 94-06)
Description of amendment request: The proposed amendments would
revise the combined Technical Specifications (TS) for the Diablo Canyon
Power Plant (DCPP) Unit Nos. 1 and 2 to change TS 3/4.1.2.5, ``Borated
Water Sources - Shutdown,'' TS 3/4.1.2.6, ``Borated Water Sources -
Operating,'' and TS 3/4.5.5, ``Emergency Core Cooling Systems -
Refueling Water Storage Tank.'' The changes delete the minimum
refueling water storage tank (RWST) solution temperature and increase
the allowed outage time (AOT) of the RWST for adjustment of boron
concentration from 1 hour to 8 hours. Specifically, the minimum RWST
temperature requirement of TS 3.1.2.5b(3), TS 4.1.2.5b, and TS 4.5.5b
would be deleted. TS 3/4.1.2.6 would be revised as follows: (1) TS
3.1.2.6b, Action Statement b., and TS 4.1.2.6b, pertaining to the RWST,
would be deleted. (2) Editorial changes would be made to reflect the
deletion of the RWST requirements. TS 3/4.5.5 would be revised as
follows: the minimum RWST temperature requirement of TS 3.5.5c would be
deleted, and the action statement would be deleted and replaced with
two action statements. Action Statement a. would specify the
requirements when the RWST is inoperable due to boron concentration.
The action statement would also provide 8 hours to restore the boron
concentration to within the required limits. If boron concentration is
not restored within 8 hours, the action statement requires that the
unit be in hot standby within 6 hours and in cold shutdown within the
following 30 hours. Action Statement b. would specify the requirements
when the RWST is inoperable due to reasons other than boron
concentration. The associated Bases would also be appropriately
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:
a. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The increase in the RWST AOT does not alter the plant
configuration or operation. The potential for the RWST boron
concentration to be outside the TS limits is small because the RWST
and its contents are not involved with normal plant operation and
are not subject to process variations associated with plant
operation.
The potential causes of boron concentration deviation have been
evaluated with the conclusion that any deviation in RWST boron
concentration would not be expected to increase significantly during
the proposed 7 hour AOT increase.
The increase in the RWST AOT from 1 hour to 8 hours for reasons
directly related to boron concentration does not have a significant
effect on the accident analyses.
The removal of the redundant statement of RWST requirements from
TS 3.1.2.6 is an administrative change with no impact on plant
operation.
The removal of the minimum temperature limit for the RWST has no
effect on the plant configuration or operation. The removal of the
temperature limits does not affect any accident analyses since
evaluations have demonstrated that, due to the moderate climate at
DCPP, the RWST will not exceed the limits assumed in DCPP accident
analyses.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
b. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Increasing the RWST AOT from 1 hour to 8 hours for reasons
directly related to boron concentration does not require physical
alteration to any plant system and does not change the method by
which any safety-related system performs its function.
The removal of the redundant statement of RWST requirements from
TS 3.1.2.6 is an administrative change that does not affect the
design and operation of the plant.
Deletion of the RWST temperature has no impact on any accident
analysis due to the moderate climate at DCPP. Additionally, the
deletion of the temperature does not require any physical alteration
to the plant or change the method by which any safety-related system
performs its function.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
c. Does the change involve a significant reduction in a margin
of safety?
Increasing the RWST AOT for reasons directly related to boron
concentration does not affect any accident analysis assumptions,
initial conditions, or results. The margins of safety reflected in
the DCPP TS are not compromised by the 7 hour AOT increase.
Consequently, the proposed change does not have an effect on margin
of safety.
The removal of the redundant statement of RWST requirements from
TS 3.1.2.6 is an administrative change that does not affect the
requirements for the RWST nor alter its function.
The removal of the RWST temperature limits will not affect the
assumptions of any accident analysis because the moderate climate at
DCPP will prevent the temperature assumptions in the analyses from
being exceeded.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment 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 &
Electric Company, P.O. Box 7442, San Francisco, California 94120
NRC Project Director: Theodore R. Quay

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 amendment request: August 17, 1994 (Reference LAR 94-07)
Description of amendment request: The proposed amendments would
revise the combined Technical Specifications (TS) for the Diablo Canyon
Power Plant Unit Nos. 1 and 2 to relocate TS 3/4.4.2.1, ``Safety Valves
- Shutdown,'' 3/4.4.7, ``Chemistry,'' 3/4.4.9.2, ``Pressurizer
(Temperature Limits),'' 3/4.4.10, ``Structural Integrity,'' and 3/
4.4.11, ``Reactor Vessel Head Vents,'' in accordance with the
Commission's Final Policy Statement for relocation of current TS that
do not satisfy any of the screening criteria for retention. As part of
the relocation of TS 3/4.4.2.1, TS 3/4.4.2.2, ``Safety Valves -
Operating,'' would be revised to require that the pressurizer safety
valves be operable in Mode 4 with the reactor coolant system cold-leg
temperature greater than the low-temperature overpressure protection
system enable temperature, and TS 6.8, ``Procedures and Programs,''
would be revised to include the reactor coolant pump flywheel
inspection program. The specific TS changes proposed are as follows:
(1)
Technical Specifications (TS) 3/4.4.2.1 ``Safety Valves -
Shutdown,'' 3/4.4.7, ``Chemistry,'' 3/4.4.9.2, ``Pressurizer
(Temperature Limits),'' 3/4.4.10, ``Structural Integrity,'' 3/4.4.11,
``Reactor Vessel Head Vents,'' and TS 6.8, ``Procedures and Programs,''
would be revised in accordance with the Commission's Final Policy
Statement on TS Improvements for Nuclear Power Reactors.
(2)
TS 3/4.4.2.2, ``Safety Valves - Operating,'' would be revised to
require that the pressurizer safety valves be operable in Mode 4 with
the reactor coolant system cold-leg temperature greater than the low-
temperature overpressure protection system enable temperature.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
a. Do the changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes simplify the TS, meet regulatory
requirements for relocated TS, and implement the recommendations of
the Commission's Final Policy Statement on TS Improvements. Future
changes to these requirements will be controlled by 10 CFR 50.59.
The proposed changes are administrative in nature and do not involve
any modifications to any plant equipment or affect plant operation.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
b. Do the changes create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes are administrative in nature, do not
involve any physical alterations to any plant equipment, and cause
no change in the method by which any safety-related system performs
its function.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
c. Do the changes involve a significant reduction in a margin of
safety?
The proposed changes do not alter the basic regulatory requirements
and do not affect any safety analyses. Therefore, the proposed changes
do not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment 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 &
Electric Company, P.O. Box 7442, San Francisco, California 94120
NRC Project Director: Theodore R. Quay

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 amendment request: August 17, 1994 (Reference LAR 94-09)
Description of amendment request: The proposed amendments would
revise the combined Technical Specifications (TS) for the Diablo Canyon
Power Plant (DCPP) Unit Nos. 1 and 2 to change TS 3/4.4.9.1, ``Reactor
Coolant System - Pressure/Temperature Limits,'' Figures 3.4-2,
``Reactor Coolant System Heatup Limitations - Applicable Up to 8
EFPY,'' and 3.4-3, ``Reactor Coolant System Cooldown Limitations -
Applicable Up to 8 EFPY,'' to extend the applicability up to 12
effective full-power years (EFPYs). TS 3/4.4.9.3, ``Overpressure
Protection Systems,'' would be revised to specify a new low-temperature
overprotection (LTOP) system actuation pressure setpoint. The
associated Bases would also be appropriately revised. Additionally, TS
3/4.1.2.2, ``Flow Paths - Operating,'' TS 3/4.1.2.4, ``Charging Pumps -
Operating,'' TS 3/4.4.1.3, ``Hot Shutdown,'' TS 3/4.4.1.4.1, ``Cold
Shutdown - Loops Filled,'' TS 3/4.4.9.3, and TS 3/4.5.3, ``Tavg Less
than 350 Degrees F,'' would be revised to specify a new LTOP system
enable temperature.
(1) In TS 3/4.4.9.1, Figure 3.4-2, ``Reactor Coolant System Heatup
Limitations - Applicable Up to 8 EFPY,'' and Figure 3.4-3, ``Reactor
Coolant System Cooldown Limitations - Applicable Up to 8 EFPY,'' are
revised as follows:
(a) The ``Controlling Materials'' for the pressure/temperature
curves are revised to reflect the current reactor vessel beltline
region limiting weld and plate materials. (b)The title for the figures
is changed to reflect the applicability of the pressure/temperature
curves for up to 12 EFPYs of service life.
(2) The proposed changes to TS 3/4.4.9.3 are as follows:(a) The
LTOP enable temperature would be changed from 323 deg.F to 270 deg.F to
be consistent with Branch Technical Position (BTP) RSB 5-2, Revision 1,
Branch Position B.2.
(b) LTOP system actuation pressure setpoint would be revised from
less than or equal to 450 psig to less than or equal to 435 psig.
(3) TS 3/4.1.2.2, TS 3/4.1.2.4, TS 3/4.4.1.3, TS 3/4.4.1.4.1, TS 3/
4.4.9.3, and TS 3/4.5.3 would be revised to change the LTOP enable
temperature from 323 deg.F to 270 deg.F to be consistent with BTP RSB
5-2, Revision 1, Branch Position B.2. TS Bases 3/4.4.9.1 would be
revised to delete a reference to Table 4.4-5, ``Reactor Vessel Material
Surveillance Program - Withdrawal Schedule.'' The table was deleted
from the TS in Amendments 54 and 53 issued in July 1990. Reference to
the table in Bases 3/4.4.9 was inadvertently not deleted. The
information in this table is currently contained in the Final Safety
Analysis Report (FSAR) Update.
(4) TS Bases 3/4.4.9.3 would be revised to discuss limitations on
reactor coolant pump (RCP) and emergency core cooling system/chemical
and volume control system pump operation during low reactor coolant
system (RCS) temperature conditions.
(5) The other affected TS Bases would also be 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:
a. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes to Figures 3.4-2 and 3.4-3 of TS 3.4.9.1
and the associated Bases will extend the applicability of the RCS
heatup and cooldown pressure/temperature limits from 8 to 12 EFPY.
Since the level of reactor vessel embrittlement projected for 12
EFPY is bounded by that previously projected for 8 EFPY, the
proposed changes will not impact the probability of brittle fracture
of the vessel, and consequently not impact the consequences of an
accident.
The present LTOP pressure setpoint was reviewed and found to be
acceptable and conservative for the extension of the pressure/
temperature curves to 12 EFPY. However, as a result of issues
unrelated to the change in the applicability of the pressure/
temperature curves, the LTOP actuation pressure setpoint is reduced.
The change accounts for pressure measurement error identified in NRC
IN [Information Notice] 93-58, a time delay in the LTOP system
actuation introduced as part of the installation of the Eagle 21
protection system, and additional conservatism incorporated into the
DCPP LTOP analysis. The changes to the pressure setpoint are
conservative and provide assurance that the maximum cold RCS
pressure will not be exceeded.
The proposed change to TS 3/4.1.2.2, 3/4.1.2.4, 3/4.4.1.3, 3/
4.4.9.3, and 3/4.5.3 will revise the LTOP enable temperature to be
consistent with the methodology and definition of ``low
temperature'' provided in BTP RSB 5-2 Revision 1. The proposed
changes do not involve physical alteration of the LTOP system or
change the method by which the LTOP system performs its function.
The proposed changes will benefit DCPP by expanding the RCS
pressure/temperature window, thereby increasing operator flexibility
during heatup and cooldown. This will decrease the probability of an
accident by decreasing the likelihood of an inadvertent PORV [power-
operated relief valve] actuation.
Deletion of reference to Table 4.4-5 from TS Bases 3/4.4.9 is
administrative in nature and does not affect plant operation.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
b. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes to TS 3.4.9.1 do not involve any physical
alteration to any plant system or change the method by which any
safety-related system performs its function. The probability of
catastrophic failure of the reactor vessel will not be changed as a
result of the extension of the curves to 12 EFPY.
The present LTOP pressure setpoint was reviewed and found to be
acceptable and conservative for the extension of the pressure/
temperature curves to 12 EFPY. However, as a result of issues
unrelated to the change in the applicability of the pressure/
temperature curves, the LTOP actuation pressure setpoint is reduced.
The change accounts for pressure measurement error identified in IN
93-58, a time delay in the LTOP system actuation introduced as part
of the installation of the Eagle 21 protection system, and
additional conservatism incorporated into the DCPP LTOP analysis.
The changes to the pressure setpoint are conservative and provide
assurance that the maximum cold RCS pressure will not be exceeded.
The proposed change to TS 3/4.1.2.2, 3/4.1.2.4, 3/4.4.1.3, 3/
4.4.9.3, and 3/4.5.3 will revise the LTOP enable temperature to be
consistent with the methodology and definitions provided in BTP RSB
5-2, Revision 1. Additionally, the proposed changes will not affect
the ability of the LTOP system to provide pressure relief at low
temperatures, thereby maintaining the LTOP design basis.
Deletion of reference to Table 4.4-5 from TS Bases 3/4.4.9 is
administrative in nature and does not result in physical alterations
or changes to the operation of the plant.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
c. Does the change involve a significant reduction in a margin
of safety?
The proposed changes to TS 3.4.9.1 will extend the applicability
of the RCS heatup and cooldown pressure/temperature limits to 12
EFPY, but will not physically change these limits. The pressure/
temperature limits have been determined in accordance with 10 CFR
50, Appendix G, and include the safety margins with regard to
brittle fracture required by the ASME Code, Section III, Appendix G.
The RTndts determined for the reactor vessels at 12 EFPY are
lower than the values previously determined at 8 EFPY. Therefore,
there will be additional safety margin in the pressure/temperature
limits with respect to Appendix G requirements.
The change in the LTOP pressure setpoint is conservative and
provides assurance that the current margin of safety is maintained.
The proposed change to TS 3/4.1.2.2, 3/4.1.2.4, 3/4.4.1.3, 3/
4.4.9.3, and 3/4.5.3, will revise the LTOP enable temperature to be
consistent with the methodology and definitions provided in BTP RSB
5-2, Revision 1, which provides the requirements for reactor vessel
overpressurization protection at low temperatures.
Deletion of reference to Table 4.4-5 from TS Bases 3/4.4.9 is an
administrative change and does not involve any physical alteration
to the plant.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment 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 &
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: August 12, 1994
Description of amendment request: The amendment would revise the
Limiting Condition for Operation for the Emergency Core Cooling System
specified in Technical Specifications Section 3.5.1 and associated
Bases Section 3.4.5.1 to include a new ACTION statement in the event
that the High Pressure Coolant Injection system and one Core Spray
subsystem, and/or one Low Pressure Coolant Injection subsystem, are
inoperable.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed Technical Specifications (TS) change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The proposed TS change does not involve any physical changes to
plant systems or components, nor does it affect the ability of the
Low pressure Coolant Injection (LPCI) Core Spray (CS), and High
Pressure Coolant Injection (HPCI) systems to respond to an accident.
These systems are not accident initiators, since their design
function is accident mitigation.
This proposed TS change, which only addresses equipment status,
will not significantly increase the probability of occurrence of an
accident previously evaluated. The addition of the proposed ACTION
statement enables the plant not to implement TS Section 3.0.3, which
requires a plant shutdown, when the HPCI system is inoperable in
conjunction with one (1) CS subsystem, and/or one (1) LPCI
subsystem. The proposed TS change does not impact the operation of
any equipment important to safety. This proposed TS change does not
make physical modifications to the plant or to equipment, nor does
it impact any design requirements of the HPCI, CS, and LPCI systems.
The proposed TS change does not introduce any failure mechanisms of
a different type than those previously evaluated, since no physical
changes are being made to the facility. This proposed change will
not create any new failure modes which would cause plant equipment
to malfunction more frequently than previously evaluated.
The basis for TS Sections 3.8.2.1 and 3.8.3.1, which specify
that four (4) independent divisions of Safeguard dc electrical power
shall be operable, or shall be restored to operability with 8 hours,
is to ensure that sufficient power is available to supply safety-
related equipment required to safely shut down the plant, and to
provide for mitigation and control of accident conditions at the
plant. As discussed in Section 6.3.2 of the NRC Safety Evaluation
Report (SER), i.e., NUREG-0991, ``Safety Evaluation Report Related
to the Operation of Limerick Generating Station, Units 1 and 2,''
dated August 1983, the most limiting single failure for the
Emergency Core Cooling System (ECCS), which includes all break
sizes, is the failure of the dc power system common to the HPCI
system, one (1) CS subsystem, and one (1) LPCI subsystem. Only one
(1) single failure is assumed to occur in the event of a Design
Basis Accident (DBA). Therefore, three (3) LPCI pumps, one (1) CS
subsystem, and the Automatic Depressurization (ADS) system would be
operable and available, for use in the event of a DBA, to provide
sufficient core cooling to safely shut down the plant. Although the
loss of Division 2 dc power specifically impacts the ``B'' LPCI and
``B'' CS, the analysis performed in the NRC SER evaluates the number
of ECCS available for use in a DBA. Since the amount of available
core cooling is independent of which loop of LPCI or CS is assumed
to fail, this analysis is applicable to the loss of any division/
loop of LPCI or CS. Therefore, the loss of the HPCI system, one (1)
CS subsystem, and/or one (1) LPCI subsystem is bounded by the
existing analysis. Since the loss of HPCI, one (1) CS subsystem,
and/or one (1) LPCI subsystem is an analyzed condition, and actions
associated with TS Section 3.0.3 are related to unanalyzed
conditions, the requirements of TS Section 3.0.3 are not applicable
to this scenario. Adding an ACTION statement, as proposed, identical
to the ACTION statement which currently applies to the loss of
Division 2 of Safeguard dc electrical power causes no change in the
consequences of any accidents previously evaluated. This proposed TS
change does not impact systems, structures, and components designed
to mitigate the consequences of an accident. In the event of an
accident, the plant configuration following the event will be within
the bounds of the existing analysis, and there will be no change in
the radiological consequences due to an accident.
Therefore, the proposed TS change does not involve an increase
in the probability or consequences of an accident [previously]
evaluated.
2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed TS change does not require any physical changes to
plant systems or equipment, nor will it affect the ability of the
HPCI, CS, and LPCI systems from performing their design functions,
which is to mitigate the consequences of an accident. These systems
do not contribute to the initiation of an accident, since their
function is accident mitigation. This proposed TS change will not
introduce new equipment malfunction or failure modes. The proposed
TS change will not introduce any failure mechanisms of a different
type than those previously evaluated. The existing design basis for
the plant, as described in Section 6.3.2.5 of the LGS Updated Safety
Analysis Report (UFSAR) and Section 6.3.2 of the NRC SER, bounds the
condition proposed by this TS Change Request. Section 6.3.2 of the
NRC SER indicates that the most limiting single failure for the ECCS
is the loss of the dc system powering the HPCI, CS, and LPCI
systems. Assuming this failure, three (3) LPCI pumps, one (1) CS
subsystem, and the ADS would still be operable and available, for
use in the event of a DBA, to ensure adequate core cooling to safely
shut down the plant. Although the loss of Division 2 dc power
specifically affects ``B'' LPCI and ``B'' CS, the analysis performed
in the NRC SER evaluates the number of ECCS available for use in a
DBA. Since the amount of available core cooling is independent of
which loop of LPCI or CS is assumed to fail, this analysis is
applicable to the loss of any division/loop of LPCI or CS. Since the
loss of HPCI, one (1) CS subsystem, and/or one (1) LPCI subsystem,
is an analyzed condition, and the actions associated with TS Section
3.0.3 pertain to unanalyzed conditions, the requirements of TS
Section 3.0.3 do not apply to the condition proposed by this TS
Change Request.
Therefore, the proposed TS change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed TS change does not involve a significant
reduction in a margin of safety.
The proposed TS change [TS] does not involve any physical
changes to the design or functional requirements of the LPCI, CS, or
HPCI systems. These systems will continue to function as designed to
mitigate the consequences of an accident.
This proposed TS change involves adding an additional ACTION
statement, and revising the associated supporting Bases section, to
specifically address the inoperability of the HPCI system in
conjunction with the inoperability of one (1) CS subsystem, and/or
one (1) LPCI subsystem. These systems would be inoperable in the
event of the loss of Division 2 of the Safeguard dc electrical power
supply. The Bases associated with Safeguard electrical power
systems, which provide power to equipment required to safely
shutdown the plant and to mitigate consequences of an accident, are
unchanged. The proposed TS change involves adding an ACTION
statement which is identical to the ACTION statement which addresses
the inoperability of Division 2 of Safeguard dc power, which is a
condition analyzed in the LGS UFSAR and NRC SER. Therefore, the
proposed TS change to include an additional ACTION statement does
not involve a reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101
NRC Project Director: Mohan C. Thadani, Acting

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania

Date of amendment request: August 23, 1994
Description of amendment request: This amendment would remove the
125/250 Vdc Class 1E Battery Load Cycle Table from Technical
Specifications, which is consistent with NUREG-1433, ``Standard
Technical Specifications.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed TS change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
This proposed change removes the repetitious 125/250 Vdc Class
1E Battery Load Cycle Table which is also found in the LGS Updated
Final Safety Analysis Report (UFSAR). The proposed change to TS does
not affect the requirement to perform surveillance testing and the
manner of performing surveillance testing is adequately described in
plant procedures. The UFSAR containing the Battery Load Cycle Table
and station procedures are maintained using the provisions of 10 CFR
50.59 and are subject to the change control process in the
Administrative Controls Section of the LGS TS Section 6.0. Since any
future changes to these controlled documents will be evaluated per
10 CFR 50.59, no [changes] (significant or insignificant) in the
probability or consequences of an accident previously evaluated will
be allowed. Therefore, this change will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
This proposed change removes the repetitious 125/250 Vdc Class
1E Battery Load Cycle Table which is also found in the LGS Updated
Final Safety Analysis Report (UFSAR). This change will not alter the
plant configuration (no new or different type of equipment will be
installed) or make changes to methods governing normal plant
operations. This change will not impose different requirements and
adequate control of information will be maintained. The manner of
performing surveillance testing can be adequately described in plant
procedures. The proposed change will remove the table, and will not
alter assumptions made in the safety analysis and licensing basis.
Therefore, this change will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed TS change does not involve a significant
reduction in a margin of safety.
This proposed change removes the repetitious 125/250 Vdc Class
1E Battery Load Cycle Table which is also found in the LGS Updated
Final Safety Analysis Report (UFSAR). The change will not reduce the
margin of safety since the location of the Battery Table has no
impact on any safety analysis assumptions. Since all Battery Load
Table changes (i.e., UFSAR Changes) and procedure changes are
evaluated per the requirements of 10 CFR 50.59, no reduction
(significant or insignificant) in the margin of safety will be
allowed. Therefore, this change will not involve a significant
reduction in a margin of safety.
The existing requirements for NRC review and approval of
revisions, in accordance with 10 CFR 50.90, to those details and
requirements proposed for deletion, do not have a specific margin of
safety upon which to evaluate. However, since the proposed change is
consistent with the BWR Standard Technical Specifications (NUREG-
1433), revising the TS to reflect the approved level of detail and
requirements ensures 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: 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: Mohan C. Thadani, Acting

Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey

Date of amendment request: August 5, 1994
Description of amendment request: The proposed amendment
incorporates line item Technical Specification improvements listed in
Generic Letter 93-05 relevant to Emergency Diesel Generator (EDG)
surveillance requirements. The proposed amendment eliminates the
requirements to start EDGs with an inoperable offsite circuit(s) of AC
electrical power and adds a provision that eliminates required testing
of the remaining EDGs when one EDG is inoperable due to an inoperable
support system or an independently testable component with no potential
for common mode failure for the remaining EDGs. In addition, if testing
of the EDGs is required, then the surveillances will be performed
within 16 hours instead of 24 hours as currently specified.
The proposed amendment also deletes the requirement to perform a
loss of offsite power (LOP) test following the 24-hour EDG endurance
run test. In its place, a hot restart test (no LOP load sequencing)
will be established.
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 not involve a significant increase in the probability or
consequences of an accident previously evaluated.
LCR 94-10
The proposed changes in this License Change Request (LCR) have
been extensively reviewed by the NRC during the preparation of
NUREG-1366 and Generic Letter 93-05, and by [Public Service Electric
and Gas Company] PSE&G during the development and approval of this
LCR. The LCR revises the current ACTION statement of Technical
Specification 3.8.1.1 to eliminate testing of the unaffected
Emergency Diesel Generators (EDGs) upon loss of an offsite power
circuit(s) and/or an EDG. The basis for this testing was originally
to verify the reliability of the EDGs, however, as stated in NUREG-
1366, industry experience has shown that excessive testing of the
EDGs has in fact reduced reliability.
The EDG design and function remain as previously analyzed and
the EDG response during accident conditions is not affected. This
change will improve EDG performance by reducing the number of
unnecessary starts and by requiring more appropriate testing (within
16 hours instead of 24 hours) when there is a potential common mode
failure.
These changes will not result in a significant increase in the
probability or consequences of a previously evaluated accident, nor
will it result in a significant reduction in a margin of safety.
LCR 94-13
The proposed changes in this License Change Request (LCR) have
been extensively reviewed by the NRC during the preparation of
NUREG-1366 and Generic Letter 93-05, and by PSE&G during the
development and approval of this LCR. Regulatory Guide 1.108, Rev.
1, states that the performance of a loss of Off-site Power (LOP)
test (Surveillance Requirement 4.8.1.1.2.h.4.b) immediately
following the 24 hour endurance run demonstrates that the Emergency
Diesel Generator (EDG) can start in the prescribed time when the EDG
is at its normal operating temperature. The purpose of performing
the LOP test immediately following the 24 hour endurance run is to
demonstrate the hot restart capability of the EDG at full load
conditions. However, demonstrating diesel generator hot restart
capability without loading the engine does not invalidate or reduce
the effectiveness of the hot restart test. Performance of this test
can be conducted in any plant condition since its performance at
power will have no adverse effect on plant operations.
The LOP test will continue to be performed at standby conditions
to provide assurance that the EDG is capable of responding to a LOP
as assumed in the accident analyses.
EDG design and function remain as previously analyzed. Their
response during accident conditions [is] not affected by these
changes. Therefore, no significant increase in the probability of an
accident previously evaluated results from these changes.
2. Will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
LCR 94-10
The elimination of the unnecessary EDG starts will not result in
any change in plant configuration or operation. Therefore, the
proposed changes will not create the possibility of a new or
different kind of accident from any accident previously evaluated or
analyzed.
LCR 94-13
The proposed revisions to the Technical Specifications do not
involve a physical change in any system configuration and do not
introduce new operating configurations. These changes will not
result in any net reduction in testing and will not affect EDG
reliability. This test may be performed in any plant condition since
its performance at power will have no adverse effect on plant
operations. Therefore, these changes do not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Will not involve a significant reduction in a margin of
safety.
LCR 94-10
The changes proposed in this LCR do not reduce the ability of
any system or component to perform its safety related function. The
basis of NUREG-1366, Generic Letter 93-05 and the analysis performed
in support of this LCR is that the reduction in unnecessary EDG
starts can improve safety by diminishing challenges to plant systems
and reducing equipment wear or degradation. These proposed changes
involve only surveillance frequencies and do not change the method
of performing any surveillance. The operation of systems and
equipment remains unchanged. Therefore, eliminating unnecessary EDG
starts does not involve a reduction in the margin of safety.
LCR 94-13
Surveillance testing per the proposed Technical Specifications
would continue to demonstrate the ability of the EDGs to perform
their intended function of providing electrical power to the
emergency safety systems needed to mitigate design basis transients
consistent with the plant safety analyses. The margin of safety
demonstrated by the plant safety analyses is therefore not affected
by the proposed changes.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: Mohan C. Thadani, Acting

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey

Date of amendment request: August 19, 1994
Description of amendment request: The proposed changes add a new
statement (b) to Limiting Condition for Operation (LCO) 3.1.3.2.1, Rod
Position Indication Systems, and reletters the existing action
statement (b) to (c). The new action (b) will read:
With two or more analog rod position indicators per bank
inoperable, within one hour restore the inoperable rod position
indicator(s) to OPERABLE status or be in Hot Standby within the next
6 hours. A maximum of one rod position indicator per bank may remain
inoperable following the one hour, with Action (a) above being
applicable from the original entry time into the LCO.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The request (both proposed changes) does not change any
assumption or parameter assumed to function in any of the design/
licensing basis analysis, and therefore the probability or
consequences of an accident previously evaluated are not increased.
The change, as described in section IB, [the addition of the new
action statement] incorporates into the applicable LCO the action
statement which is already taken under technical specification
3.0.3, and does not alter the operator response or response time.
2. Does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
The proposed change does not introduce any design or physical
configuration changes to the facility which could create new
accident scenarios.
3. Does not involve a significant reduction in a margin of
safety.
As stated in response to question number 1 above, the change
does not change any assumption or parameter assumed to function in
any of the design/licensing basis analysis. No changes to the
operator response or operator response time is proposed, only that
the response is now taken under the confines of the LCO.
Therefore, there is no reduction in any margin of safety from
the proposed changes.
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: Salem Free Public library, 112
West Broadway, Salem, New Jersey 08079
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
NRC Project Director: Mohan C. Thadani, Acting

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: July 20, 1994, as supplemented September
20, 1994
Description of amendment request: The proposed change would modify
the Virgil C. Summer Nuclear Station, Unit 1, (VCSNS) Technical
Specifications (TS) to allow alternative, equivalent testing of diesel
fuel used in the emergency diesel generators (EDG). These alternative
methods are necessary due to recent changes in Environmental Protection
Agency (EPA) Regulations that are designed to limit the use of high
sulfur fuels. The licensee also proposes to modify the VCSNS TS by
changing the revision level of WCAP-10216-P-A, ``Relaxation of Constant
Axial Offset Control - FQ Surveillance Technical Specification,''
referenced in TS 6.9.1.11. This pertains to the FQ(z) TS (TS 3.2.1 and
3.2.2) and is necessary since Westinghouse revised their methodology in
determining FQ(z).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The probability or consequences of an accident previously
evaluated is not significantly increased.
The change in testing methods for the EDG fuel oil has no impact
on the probability or consequences of any design basis accident.
These tests have been determined to be equivalent to the previously
approved testing methods and are needed due to changes in the EPA's
regulations regarding sulfur in motor vehicle fuels. The dye used to
identify high sulfur fuels will have no adverse affect on the
performance of the EDG's. The proposed testing assures a continued
high level of quality of the diesel fuel received and stored on
site.
The change in revision level of a reference in TS section
6.9.1.11 has no impact on the probability of occurrence or
consequences of any design basis accident. All design and
performance criteria will continue to be met and no new single
failure mechanisms will be created. The change in revision level for
WCAP-10216-P-A does not involve any alterations to plant equipment
or procedures which could affect any operational modes or accident
precursors. This change only incorporates by reference, the
methodology for determining the penalty to be used in calculating
Core Operating Limits. This methodology allows the penalty to be
cycle specific and is primarily affected by the core configuration.
This penalty is used for normal operation and provides more
conservatism to the core operation for the cycle.
2. [The proposed license amendment does not] create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The change in testing methods for the EDG fuel oil will not
create the possibility of a new or different kind of accident from
any accident previously evaluated. These tests have been determined
by the EPA and other organizations to be equivalent to the
previously approved testing methods. The effect of the blue dye,
used to identify high sulfur fuels, on the performance of the EDGs
has been evaluated and determined to be insignificant. The testing
proposed assures a continued high level of quality for the diesel
fuel received and stored on site.
The change of revision level of a reference in TS section
6.9.1.11 has no impact on the probability of occurrence or
consequences of any design basis accident. All design and
performance criteria will continue to be met and no new single
failure mechanisms will be created. The change in revision level for
WCAP-10216-P-A does not involve any alterations to plant equipment
or procedures which could affect any operational modes or accident
precursors. This change only incorporates, by reference, the
methodology for determining the penalty to be used in calculating
Core Operating Limits. This methodology allows the penalty to be
cycle specific and is primarily affected by the core configuration.
This penalty is used for normal operation and provides more
conservatism to the core operation for the cycle.
3. [The proposed license amendment does not] involve a
significant reduction in a margin of safety.
The change in testing methods for the EDG fuel oil will not
involve a significant reduction in a margin of safety. The proposed
testing methods have been determined to be equivalent to the
previously approved testing methods. The test for sulfur assures
that the sulfur content is within the allowable range for weight-
percent. The test for color and clarity assures that the fuel is
relatively free of water and particulate contaminants. The proposed
tests provide at least an equivalent level of quality and
repeatability for the fuel oil analysis, thus assuring that the
margin of safety is not reduced.
The change in revision level of a reference in TS section
6.9.1.11 does not change the proposed reload design or safety
analysis limits for each cycle reload core. The associated change to
WCAP-10216-P-A due to the revision will be specifically evaluated
using approved reload design methods. The larger penalty actually
provides for an increase in margin during certain burnup ranges.
Since the safety analysis limits are unaffected, and the cycle
specific analysis will show that the analysis limits are met, the
change proposed will have no adverse impact on 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: 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: David B. Matthews

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: July 28, 1994
Description of amendment requests: The licensee proposes revisions
to Technical Specification (TS) 3.9.8.1, ``Shutdown Cooling and Coolant
Circulation -- High Water Level,'' TS 3.9.8.2, ``Shutdown Cooling and
Coolant Circulation -- Low Water Level,'' and their Bases to facilitate
testing of low-pressure safety injection system components and permit
additional flexibility in scheduling maintenance on the shutdown
cooling system.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. 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
Limiting Conditions for Operation (LCO) in Technical
Specifications (TSs) 3.9.8.1 and 3.9.8.2 define the operability
requirements for the Shutdown Cooling (SDC) system during refueling
operations (Mode 6) while the water level above the top of the
reactor vessel flange is at least 23 feet and less than 23 feet,
respectively. The objective of these TSs is to ensure that (1)
sufficient cooling is available to remove decay heat, (2) the water
in the reactor vessel is maintained below 140 degrees Fahrenheit,
and (3) sufficient coolant circulation is maintained in the reactor
core to minimize boron stratification leading to a boron dilution
incident.
The proposed TS changes affect the current limits imposed while
ensuring adherence to the basis of the TS. No plant modifications
are being made. The reactor cavity water level limitations and SDC
system required operating times are being changed based on plant
specific calculations and the objectives of the TSs are being
maintained.
(1) reduce the water level where two trains of SDC are required
from 23 feet to 20 feet above the reactor pressure vessel flange,
In the Bases Section 3/4.9.8, it is stated that ``With the
reactor vessel head removed and 23 feet of water above the reactor
pressure vessel flange, a large heat sink is available for core
cooling, thus in the event of a failure of the operating shutdown
cooling train, adequate time is provided to initiate emergency
procedures to cool the core.''
In the Bases for the New Standard Technical Specifications,
``NUREG 1432, Revision 0, dated September 30, 1992, Section B 3.9.4
it is stated that; ``The 23 ft level was selected because it
corresponds to the 23 ft requirement established for fuel movement
in LCO 3.9.6, ``Refueling Water Level.''
Southern California Edison (Edison) calculations show that there
is a minimal difference in the time to boil due to the 3-foot change
in required water level. Therefore, adequate water is still
available to mitigate the consequences of losing SDC.
(2) increase the time a required train of the SDC system may be
removed from service from up to 1 hour per 8-hour period to up to 2
hours per 8-hour period,
(3) allow the SDC system to be removed from service to allow
testing of Low Pressure Safety Injection system components,
The proposed TS changes the time the SDC train may be removed
from operation from up to 1 hour per 8-hour period to up to 2 hours
per 8-hour period, and allows removal of the SDC train from
operation for testing of the Low Pressure Safety Injection (LPSI)
system components as well as for core alterations in the vicinity of
the hot legs. The proposed TS change also imposes certain
restrictions to ensure operating the SDC system in accordance with
this proposed TS change is of no safety significance. These
restrictions are discussed separately below.
When securing the only operating train of the SDC system, the
maximum Reactor Coolant System (RCS) temperature is maintained less
than or equal to 140 degrees Fahrenheit. The initial conditions and
heatup rate are selected such that the RCS temperature remains less
than or equal to 140 degrees Fahrenheit during the test. Therefore,
there is ample margin to boiling. Typical initial temperatures are
less than 100 degrees Fahrenheit.
The water being injected by the LPSI system test is cool water
from the Refueling Water Storage Tank (RWST) and will increase the
available inventory providing the heat sink by several inches. The
two hours is sufficient time to align the system to test, perform
the test, and restore the train of SDC to operation prior to
exceeding 140 degrees Fahrenheit.
No operations are permitted that would cause a reduction of the
RCS boron concentration. This minimizes the probability of an
inadvertent boron dilution event. The use of adequately borated
water for injection into the RCS during the test provides assurance
that the test itself cannot lead to a boron dilution event. When the
SDC system is operating, the minimum SDC flow rate of 2200 gpm
imposed by TS 4.9.8.1 and TS 4.9.8.2 is sufficient to ensure
complete mixing of the boron within the RCS.
The LPSI component testing is only allowed when the reactor
cavity water level is maintained greater than or equal to 20 feet
above the reactor pressure vessel flange. This level ensures an
adequate heat sink to perform the LPSI pump suction header check
valve test.
(4) allow for running 1 train of shutdown cooling with
additional requirements when the water level is less than 20 feet
but greater than 12 feet above the reactor pressure vessel flange,
(5) add an action to be taken when operating 1 train of SDC with
less than 20 feet above the reactor pressure vessel flange when the
specified requirements are not met,
In the event of a loss of SDC, the time to boil is reduced from
approximately 3.7 hours when the water level is 23 feet above the
reactor vessel flange to approximately 2.3 hours at 12 feet,
assuming the reactor has only been shutdown for 6 days. However,
this is ample time to close containment (less than 1 hour) and to
restore SDC or initiate alternative cooling (e.g., add water to the
cavity (approximately 1 hour)). Twelve feet of water above the
reactor vessel flange corresponds to 24 feet 8-7/8 inches above the
active fuel.
Requiring the reactor to be shutdown for at least 6 days to have
only one train of SDC operable when the reactor cavity level is
between 20 feet and 12 feet above the reactor pressure vessel flange
ensures that the time to boil is greater than twice the time it
would take us to establish containment closure and to commence
reactor cavity fill with the required standby equipment.
One train of SDC operating with a containment spray pump allows
for the high capacity LPSI pump to be the main standby pump capable
of filling the reactor cavity to at least 20 feet above the reactor
pressure vessel flange upon loss of SDC. The high pressure safety
injection pump will also be maintained ready to increase the water
level if needed. In support of this contingency the RWST will be
required to contain the volume of water required to raise the level
to 20 feet above the reactor vessel flange.... The reactor cavity
can be filled at a rate of approximately 4.0 inches per minute with
the LPSI pump.
If operating one train of the SDC system with less than 20 feet
of water above the reactor pressure vessel flange and any of the
required conditions are not met, requiring immediate action to
establish greater than or equal to 20 feet of water above the
reactor pressure vessel flange ensures no time is wasted trying to
restore conditions that should be used to increase the volume of
water of the heat sink. By taking action to restore the level to 20
feet above the reactor pressure vessel flange the plant will be
placed in TS 3.9.8.1, which only requires one train of SDC to be
operable. Additionally, the core will not heat up while the water
level in the reactor cavity is being raised with cool water from the
RWST. This will provide additional time to either restore the one
train of SDC or take other actions to provide core cooling.
A Probabilistic Risk Assessment (PRA), with (a) one train of the
SDC system operable with the reactor cavity water level greater than
or equal to 12 feet above the reactor pressure vessel flange, and
(b) one train of the SDC system operable with the reactor cavity
water level greater than or equal to 20 feet above the reactor
pressure vessel flange, showed that the operations in accordance
with the proposed TS would not significantly increase the
probabilities of inventory boiling and core damage.
(6) delete the obsolete reference to the implementation of DCP
2-6863 and MMP 3-6863,
This is an editorial change.
(7) delete an obsolete footnote allowing removal of both trains
of SDC with the water less than 23 feet above the reactor vessel
flange from the Unit 3 TSs.
This is an editorial change.
Therefore, proposed changes 1 through 7 do not involve a
significant increase in the probability or consequences of an
accident.
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
(1) reduce the water level where two trains of SDC are required
from 23 feet to 20 feet above the reactor pressure vessel flange,
(2) increase the time a required train of the SDC system may be
removed from service from up to 1 hour per 8-hour period to up to 2
hours per 8-hour period,
(3) allow the SDC system to be removed from service to allow
testing of Low Pressure Safety Injection system components,
(4) allow for running 1 train of shutdown cooling with
additional requirements when the water level is less than 20 feet
but greater than 12 feet above the reactor pressure vessel flange,
(5) add an action to be taken when operating 1 train of SDC with
less than 20 feet above the reactor pressure vessel flange when the
specified requirements are not met,
The Limiting Conditions for Operation (LCO) in Technical
Specifications (TSs) 3.9.8.1 and 3.9.8.2 define the operability
requirements for the SDC system during refueling operations (Mode 6)
while the water level above the top of the reactor vessel flange is
at least 23 feet and less than 23 feet, respectively. The objective
of the proposed TS changes is to ensure that the intent of the Bases
is maintained. [i.e., (1) sufficient cooling is available to remove
decay heat, (2) water in the reactor vessel is maintained below 140
degrees Fahrenheit, and (3) sufficient coolant circulation is
maintained in the reactor core to minimize boron stratification
leading to a boron dilution incident.]
The proposed TS changes affect the current limits imposed while
ensuring adherence to the basis of the TS. No plant modifications
are being made. The reactor cavity water level limitations and SDC
system required operating times are being changed based on plant
specific calculations and the objective of the TSs are being
maintained. The added requirements and action statement facilitate
safe operation.
(6) delete the obsolete reference to the implementation of DCP
2-6863 and MMP 3-6863, and
This is an editorial change.
(7) delete an obsolete footnote allowing removal of both trains
of SDC with the water less than 23 feet above the reactor vessel
flange from the Unit 3 TSs.
This is an editorial change.
Therefore, the operation of the facility in accordance with
proposed changes 1 through 7 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
Limiting Conditions for Operation (LCO) in TSs 3.9.8.1 and
3.9.8.2 define the operability requirements for the SDC system
during refueling operations (Mode 6) while the water level above the
top of the reactor vessel flange is at least 23 feet and less than
23 feet, respectively. The objective of these TSs is to ensure that
(1) sufficient cooling is available to remove decay heat, (2) the
water in the reactor vessel is maintained below 140 degrees
Fahrenheit, and (3) sufficient coolant circulation is maintained in
the reactor core to minimize boron stratification leading to a boron
dilution incident.
(1) reduce the water level where two trains of SDC are required
from 23 feet to 20 feet above the reactor pressure vessel flange,
In the Bases Section 3/4.9.8, it is stated that ``With the
reactor vessel head removed and 23 feet of water above the reactor
pressure vessel flange, a large heat sink is available for core
cooling, thus in the event of a failure of the operating shutdown
cooling train, adequate time is provided to initiate emergency
procedures to cool the core.''
In the Bases for the New Standard Technical Specifications,
``NUREG 1432, Revision 0, dated September 30, 1992, Section B 3.9.4
it is stated that: ``The 23 ft level was selected because it
corresponds to the 23 ft requirement established for fuel movement
in LCO 3.9.6, ``Refueling Water Level.''
Edison calculations show that there is a minimal difference in
the time to boil due to the 3-foot change in required water level.
Therefore, the margin of safety has not been significantly reduced.
(2) increase the time a required train of the SDC system may be
removed from service from up to 1 hour per 8-hour period to up to 2
hours per 8-hour period,
(3) allow the SDC system to be removed from service to allow
testing of Low Pressure Safety Injection system components,
The proposed TS changes the time the SDC train may be removed
from operation from up to 1 hour per 8-hour period to up to 2 hours
per 8-hour period, and allows removal of the SDC train from
operation for testing of the LPSI system components as well as for
core alterations in the vicinity of the hot legs. The proposed TS
change also imposes certain restrictions to ensure operating the SDC
system in accordance with this proposed TS change is of no safety
significance. These restrictions are discussed separately below.
When securing the only operating train of the SDC system, the
maximum RCS temperature is maintained less than or equal to 140
degrees Fahrenheit. The initial conditions and heatup rate are
selected such that RCS temperature remains less than or equal to 140
degrees Fahrenheit during the test. Therefore, there is ample margin
to boiling. Typical initial temperatures are less than 100 degrees
Fahrenheit.
The water being injected by the LPSI system test is cool water
from the RWST and will increase the available inventory providing
the heat sink by several inches. The two hours is sufficient time to
align the system to test, perform the test, and restore the train of
SDC to operation prior to exceeding 140 degrees Fahrenheit.
No operations are permitted that would cause a reduction of the
RCS boron concentration. This minimizes the probability of an
inadvertent boron dilution event. The use of adequately borated
water for injection into the RCS during the test provides assurance
that the test itself cannot lead to a boron dilution event. When the
SDC system is operating, the minimum SDC flow rate of 2200 gpm is
sufficient to ensure complete mixing of the boron within the RCS.
The LPSI component testing is only allowed when the reactor
cavity water level is maintained greater than or equal to 20 feet
above the reactor pressure vessel flange. This level ensures an
adequate heat sink to perform the LPSI pump suction header check
valve test.
The added requirements and the nature of the test provide
assurances that the water temperature will be maintained less than
140 degrees Fahrenheit and that boron stratification is prevented.
(4) allow for running 1 train of shutdown cooling with
additional requirements when the water level is less than 20 feet
but greater than 12 feet above the reactor pressure vessel flange,
(5) add an action to be taken when operating 1 train of SDC with
less than 20 feet above the reactor pressure vessel flange when the
specified requirements are not met,
In the event of a loss of SDC, the time to boil is reduced from
approximately 3.7 hours at 23 feet to approximately 2.3 hours at 12
feet, when the reactor has only been shutdown for 6 days. However,
this is ample time to close containment (less than 1 hour), and to
restore SDC or initiate alternative cooling (e.g., add water to the
cavity (approximately 1 hour)).
Requiring the reactor to be shutdown for at least 6 days to have
only one train of SDC operable when the reactor cavity level is
between 20 feet and 12 feet above the reactor pressure vessel flange
ensures that the time to boil is greater than twice the time it
would take us to establish containment closure and to commence
reactor cavity fill with the required standby equipment.
One train of SDC operating with a containment spray pump allows
for the high capacity LPSI pump to be the main standby pump capable
of filling the reactor cavity to at least 20 feet above the reactor
pressure vessel flange upon loss of SDC. The high pressure safety
injection pump will also be maintained ready to increase the water
level if needed. In support of this contingency the RWST will be
required to contain the volume of water required to raise the level
to 20 feet above the reactor vessel flange. The reactor cavity can
be filled at a rate of approximately 4.0 inches per minute with the
LPSI pump.
If operating one train of the SDC system with less than 20 feet
of water above the reactor pressure vessel flange and any of the
required conditions are not met, requiring immediate action to
establish greater than or equal to 20 feet of water above the
reactor pressure vessel flange ensures no time is wasted trying to
restore conditions that should be used to increase the volume of
water of the heat sink. By taking action to restore the level to 20
feet above the reactor pressure vessel flange the plant will be
placed in TS 3.9.8.1, which only requires one train of SDC to be
operable. Additionally, the core will not heat up while the reactor
cavity water level is being raised with cool water from the RWST.
This will provide additional time to either restore the one train of
SDC or take other actions to provide core cooling.
A PRA showed that the operations in accordance with the proposed
TS did not significantly increase the probabilities of inventory
boiling and core damage.
(6) delete the obsolete reference to the implementation of DCP
2-6863 and MMP 3-6863,
This is an editorial change.
(7) delete an obsolete footnote allowing removal of both trains
of SDC with the water less than 23 feet above the reactor vessel
flange from the Unit 3 TSs.
This is an editorial change.
Therefore, operation of the facility in accordance with proposed
changes 1 through 7 do not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment 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

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of amendment request: September 9, 1994 (TS 94-04)
Description of amendment request: The proposed change would revise
specifications associated with the cold leg accumulators (CLAs).
Specifically, the proposed amendment would: (1) remove a footnote from
Specification 3.5.1.1.c that applied to Unit 2 Cycle 6 operation only;
(2) add a requirement to Specification 3.5.1.1 that power be removed
from the CLA isolation valve when the reactor coolant system pressure
is above 2000 psig; (3) modify Specification 3.5.1.1 Action Statement
a. to indicate that with a CLA inoperable for reasons other than the
boron concentration not being within limits, the CLA must be returned
to operable status within 1 hour or the plant placed in the hot standby
condition, and the pressurizer pressure reduced to 1000 psig or less
within the next 6 hours; (4) modify Specification 3.5.1.1 Action
Statement b. to indicate that with a CLA inoperable because the boron
concentration is not within limits, the boron concentration must be
restored to within limits within 72 hours or the plant placed in the
hot standby condition within the next 6 hours and the pressurizer
pressure reduced to 1000 psig or less within the next 6 hours; (5)
remove the wording from Specification 4.5.1.1.1.a.1 for using the
absence of alarms or level measurement as the technique used to verify
CLA volume and pressure; (6) add the requirement to Specification
4.5.1.1.1.a.2 to verify that the CLA isolation valve is ``fully open''
rather than ``open;'' (7) modify Specification 4.5.1.1.1.b to show that
verification of boron concentration is not required for additions from
the refueling water storage tank, and add a footnote to indicate that
the verification is required only if the affected accumulator
experienced a volume increase; (8) modify Specification 4.5.1.1.1.c to
show that the test is satisfied by verifying that power is removed from
the isolation valve, not that the valve operator is disconnected by
removal of the breaker from the circuit; (9) delete Specification
4.5.1.1.1.d to verify that each CLA isolation valve opens automatically
when reactor coolant pressure exceeds the P-11 setpoint, and upon
receipt of a safety injection signal; (10) delete Specification
4.5.1.1.2 to verify the accumulator water level and pressure channels
operable by performing Channel Functional and Calibration tests, and
delete the related footnote; (11) change ``tanks'' to ``each cold leg
injection accumulator;'' and (12) revise the associated Bases where
necessary to reflect these changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
TVA has evaluated the proposed technical specification (TS)
change and has determined that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance
with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes to TS 3.5.1.1 implement revised action
times for cold leg injection accumulator (CLA) inoperability.
Several other clarifications and enhancements have been incorporated
to provide consistency with the latest version of standard TSs
(NUREG-1431). The new action times provide a prompt one-hour action
to initiate unit shutdown for conditions that could prevent the
injection of a CLA into the core. For boron concentration outside
limits, a 72-hour action to restore CLA concentration is allowed
because the CLA can still perform the core injection safety
function. The removal of surveillance requirements (SRs) for
verifying automatic opening features for the CLA isolation valves
does not impact the required TS alignment that is assumed in the
safety analysis. The instrumentation calibration and functional test
SRs have also been removed based on the instrumentation only
providing CLA level and pressure indications for TS compliance and
not performing an accident mitigation function. The above changes do
not alter the required limits for CLA operability or system
configurations. These changes are consistent with NUREG-1431 and
provide acceptable flexability[sic] for CLA operability verification
and surveillance testing and reasonable actions for CLA
inoperability. Since no changes have been proposed that would change
the conditions assumed for the CLAs in the accident analysis, the
consequences of an accident will not be increased. The CLAs perform
accident mitigation functions and are not considered to be the
source of an accident. Therefore, since the plant configurations and
functions are unchanged by the proposed changes, the probability of
an accident will not be increased.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed changes clarify existing CLA operability
requirements, modify action times for CLA inoperability, enhance and
simplify SRs, and remove surveillances that are not required to
verify the CLA's ability to perform safety functions. None of these
changes affect the operation of the plant or the CLA configuration
and accident mitigation capabilities. Therefore, since the CLAs will
continue to support the plant as before, these proposed changes will
not create a new or different kind of accident.
3. Involve a significant reduction in a margin of safety.
The CLA requirements for volume, pressure, boron, and isolation
valve position are not changed by the proposed request. The CLAs
will continue to provide the same safety function capabilities as
assumed in the safety analysis. Therefore, no reduction in the
margin of safety will result from these chanes because CLA functions
are unchanged.
The NRC has reviewed the licensee's analysis and, based on this
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of amendment request: September 9, 1994 (TS 94-08)
Description of amendment request: The proposed change would add
``main steam vaults'' to the footnote of Surveillance Requirement
4.6.1.1. This would allow inspection of the valves, blind flanges, and
deactivated automatic valves located in the vaults that are required to
be in the closed position during accident conditions and that are
locked, sealed, or otherwise secured in the closed position, on a cold
shutdown frequency rather than every 31 days.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
TVA has evaluated the proposed technical specification (TS)
change and has determine that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.93(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance
with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change will exempt containment isolation valves in
the east and west main steam valve vaults from examination every
thirty one days if those valves are locked, sealed or otherwise
secured. The valves and flanges that are located inside the main
steam valve vaults and are required to be closed during accident
conditions, will be verified in their required position during cold
shutdown and will be secured in this position. The environmental
conditions in these areas ensure they will be low traffic areas
where the probability of misalignment or manipulation is remote.
Loss of containment integrity is not considered to be an initiator
of any accident. This change does not affect any accident analysis
assumptions or results for SQN. Therefore, there is no increase in
the probability or consequences of an accident previously evaluated,
as a result of this change.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
This revision will not change any plant equipment, system
configurations, or accident assumptions. The appropriate components
in the valve vaults will continue to be verified in the closed
position and locked, sealed, or otherwise secured. The physical
congestion and high temperatures in the area will be effective in
maintaining this as a low traffic area that will contribute to the
low probability of misalignment or manipulation of these components
between inspections. Therefore, this change will not affect the
safety function of these components and will not create the
possibility of a new or different kind of accident.
3. Involve a significant reduction in a margin of safety.
The proposed change is consistent with current SQN accident
analysis assumptions since only the time interval between
performances of the surveillance is being extended. This change will
not impact any margin of safety.
The NRC has reviewed the licensee's analysis and, based on this
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

Date of amendment request: July 14, 1994
Description of amendment request: The proposed changes to the
Technical Specifications would remove the remaining references to
cycle-specific parameters in Technical Specification 3.12.A.2 and
associated Technical Specification Figures 3.12-1A and 1B. These
figures and the control bank insertion limits are presently specified
in the Core Operating Limits Report (COLR). The NRC-approved
methodologies presently listed in the Technical Specifications are used
to calculate and evaluate the parameter limits presented in the COLR
for each reload 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:
Specifically, operation of Surry Power Station in accordance
with the Technical Specification changes will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated. The removal of the
remaining reference to cycle-specific core operating limits and
Technical Specification Figures 3.12-1A and 1B, from the Surry
Technical Specifications has no influence or impact on the
probability or consequences of any accident previously evaluated.
The proposed amendment is administrative in nature in that it
corrects omissions from a previously approved amendment. This change
has no impact on actions to be taken when or if limits are exceeded
as is required by the current Technical Specifications. Each
accident analysis addressed in the Surry UFSAR [Updated Final Safety
Analysis Report] will be examined with respect to changes in cycle-
dependent parameters, which are determined by application of NRC-
approved reload design methodologies. The impact of these parameter
changes on transient results is then evaluated to ensure that the
results remain bounded by respective transient analysis acceptance
criteria. This examination, which is performed per the requirements
of 10 CFR 50.59, ensures that future reloads will not involve an
increase in the probability or consequences of an accident
previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated. The removal of the remaining
reference to cycle-specific core operating limits and Technical
Specification Figures 3.12-1A and 1B has no influence or impact, nor
does it contribute in any way to the probability or consequences of
any accident previously evaluated. No safety-related equipment,
safety function, or plant operating characteristic will be altered
as a result of the proposed changes. This cycle-specific variable
(control bank insertion limits) is calculated using NRC approved
methods, and the results are submitted to the NRC for information in
accordance with Technical Specification 6.2. The Technical
Specifications will continue to require operation within the
required core operating limits, and appropriate actions will be
taken when or if any of these limits are exceeded. Therefore, the
proposed amendment does not in any way create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety. The
margin of safety is not affected by this administrative change which
removes the remaining reference to cycle-specific core operating
limits and Technical Specification Figures 3.12-1A and 1B from the
Technical Specifications. The margin of safety presently provided by
current Technical Specifications remains unchanged. Appropriate
measures exist to control the values of these cycle-specific limits.
The proposed amendment continues to require operation within the
core limits which were developed from the NRC-approved reload design
methodologies. Further, the actions to be taken when or if limits
are violated remain unchanged. Development of limits for future
reloads will continue to conform to those methods described in NRC-
approved documentation. In addition, each reload requires a 10 CFR
50.59 safety review to assure that operation of the unit within the
cycle-specific limits will not involve a reduction in any margin of
safety. Therefore, the proposed changes are administrative in nature
and do not impact the operation of Surry in a manner that involves a
reduction in a margin of safety.
The NRC staff has reviewed the licens

[Text truncated at 120,000 characters. The full text is on the page linked above.]

---

Source: Frix Law Library, https://www.frixlaw.com/law-library/documents/fr%3AX94-11012. Public record. Not legal advice.