[Federal Register Volume 59, Number 196 (Wednesday, October 12, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X94-11012]


[[Page Unknown]]

[Federal Register: October 12, 1994]


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NUCLEAR REGULATORY COMMISSION
 

Biweekly Notice

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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from 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 licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Project Director: Victor M. McCree (Acting)

Virginia Electric and Power Company, Docket Nos. 50-280, 50-281, 
50-338,50-339, Surry Power Station, Units No. 1 and No. 2 Surry 
County,Virginia and North Anna Power Station, Units No. 1 and No. 
2, LouisaCounty, Virginia

    Date of amendment request: September 6, 1994
    Description of amendment request: The proposed changes would revise 
the Technical Specifications (TS) for Surry 1&2 and North Anna 1&2. 
Specifically, the proposed changes would revise the: (1) Management 
Safety Review Committee (MSRC) review responsibilities regarding safety 
evaluations and Station Nuclear Safety and Operating Committee (SNSOC) 
meeting minutes and reports, and (2) SNSOC review responsibilities for 
procedure changes. However, the changes now also state that the MSRC 
will review safety evaluations, and the SNSOC will review procedure 
changes, as programmatically discussed in the Updated Final Safety 
Analysis Report (UFSAR).
    The licensee's proposed changes revise and supersede the licensee's 
original proposed changes dated December 27, 1993 and noticed in the 
Federal Register on February 16, 1994, (59 FR 7700) for NA-1&2, and 
March 16, 1994 (59 FR 12371) for Surry 1 & 2.
    The North Anna and Surry Power Station Technical Specifications 
presently address the organization and responsibilities of both the 
onsite and offsite review groups, the SNSOC and the MSRC, respectively. 
The responsibilities of the SNSOC include the review of new procedures 
and changes to procedures that affect nuclear safety. The MSRC review 
responsibilities include the review of safety evaluations and SNSOC 
meeting minutes and reports. It is proposed that the extent of these 
review activities be revised in the Technical Specifications to ensure 
the two review groups are focusing on nuclear safety issues and not 
spending an unnecessary amount of time on administrative activities of 
minimal safety significance.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    [Specifically, operation in accordance with the proposed 
Technical Specifications changes] will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated. As administrative 
changes, the proposed Technical Specifications changes have no 
direct or indirect effect on accident precursors. No plant 
modifications are being implemented and operation of the plant is 
unchanged. SNSOC review of new procedures and procedure changes that 
require a safety evaluation ensures that activities that could 
affect nuclear safety are being properly reviewed. The MSRC's 
overview of representative samples of safety evaluations and SNSOC 
meeting minutes and reports based on performance ensures these 
programs are being properly implemented and nuclear safety is not 
being compromised; or
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated since physical modifications 
are not involved and systems and components will be operated as 
before the change. The proposed changes are wholly administrative in 
nature and have no impact on plant operations or accident 
considerations. These changes modify the scope of SNSOC review of 
procedure changes and MSRC's review functions concerning safety 
evaluations and SNSOC meeting minutes and reports. Procedure changes 
will continue to receive management review in accordance with 
administrative procedures, however, only changes that require a 
safety evaluation will require SNSOC approval. MSRC review of 
representatives samples of safety evaluations and SNSOC meeting 
minutes and reports based on performance will continue to provide 
adequate assurance that nuclear safety is being properly considered; 
or
    3. Involve a significant reduction in a margin of safety as 
defined in the basis of any Technical Specification since the 
responsibilities of the SNSOC and MSRC are not addressed by the 
existing Technical Specification Bases, nor are review requirements 
for procedures. The proposed changes are administrative in nature 
and have no impact on, nor were they considered in, existing UFSAR 
accident analyses. Safety significant procedure changes, i.e., 
changes that require a safety evaluation to be prepared, will 
continue to be reviewed by SNSOC, as will new procedures. Procedure 
changes still require cognizant management approval and preparation 
of an activity screening to determine whether or not the change 
impacts nuclear safety. This ensures activities important to nuclear 
safety are being appropriately reviewed. The effectiveness of the 
safety evaluation program, and the thoroughness of SNSOC meetings 
and reports will be assured through the MSRC's plant overview 
function which is based on observed performance.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room locations: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185, and The Alderman 
Library, Special Collections Department, University of Virginia, 
Charlottesville, Virginia 22903-2498.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Project Director: Victor McCree, Acting

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee 
Nuclear Power Plant, Kewaunee County, Wisconsin

    Date of amendment request: August 24, 1994
    Description of amendment request: The proposed amendment would 
revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS) 
3.1.b.1 and Figure TS 3.1-4 regarding Low Temperature Overpressure 
(LTOP) protection for the reactor coolant pressure boundary. Currently, 
the TS specify the LTOP requirements through the end of operating cycle 
20 or 17.14 effective full power years. The proposed change extends the 
LTOP requirements through the end of operating cycle 21 or 18.40 
effective full power years. The Basis Section would also be modified 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:
    The proposed change was reviewed in accordance with the 
provisions of 10 CFR 50.92 to show no significant hazards exist. The 
proposed change will not:
    1) involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The use of RG 1.99, Revision 2, Regulatory Position C.2 does not 
modify the reactor coolant system pressure boundary, nor make any 
physical changes to the facility design, material, construction 
standards, or setpoints. The probability of a LTOP event occurring 
is independent of the pressure temperature limits for the RCS 
pressure boundary. Therefore, the probability of a LTOP event 
occurring remains unchanged.
    The use of predicted fluence values through the end of operating 
cycle 21 is appropriately considered within the calculations in 
accordance with standard industry methodology previously docketed 
under WCAP 13227. Revised flux values were used for Cycles 16, 17, 
18 and 19 based on actual core reload designs. Previous cycles flux 
values are the same as reported in WCAP 12333.
    The calculation of pressure temperature limits in accordance 
with approved regulatory methods provides assurance that reactor 
pressure vessel fracture toughness requirements are met and the 
integrity of the RCS pressure boundary is maintained. Similar 
methodology was used in calculations to support approved amendment 
108 to the Kewaunee Technical Specifications dated April 7, 1994.
    The use of Regulatory Position C.2 and fluence values through 
EOC 21 meet previously established criteria for protection of the 
health and safety of the public. The consequences of a LTOP 
transient therefore, remain unchanged.
    2) create the possibility of a new or different type of accident 
from an accident previously evaluated.
    The use of Regulatory Position C.2 and fluence through EOC 21 
does not modify the reactor coolant system pressure boundary, nor 
make any physical changes to the LTOP setpoint or system design.
    Therefore, no new failure mechanisms are created that could 
create the possibility of an accident of a new or different type.
    3) involve a significant reduction in the margin of safety.
    The Appendix G pressure temperature limitations are calculated 
in accordance with regulatory requirements and calculational 
limitations specified in RG 1.99, Revision 2. RG 1.99, Revision 2, 
is an acceptable method for implementing the requirements of 10 CFR 
50 Appendices G and H. Similar methodology was used in calculations 
to support approved amendment 108 dated April 7, 1994. The reactor 
coolant pump starting restrictions of TS 3.1.a.1.c remain in place.
    The revised calculations meet the NRC acceptance criteria for 
the LTOP setpoint and system design as described in NRC Safety 
Evaluation Report (SER) dated September 6, 1985 which concluded that 
``the spectrum of postulated pressure transients would be 
mitigated...such that the temperature pressure limits of Appendix G 
to 10 CFR 50 are maintained.''
    The use of Regulatory Position C.2, meets previously established 
criteria for the pressure temperature limits for the LTOP system and 
setpoint. Thus, the margin of safety as described in the NRC SER is 
not reduced.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Wisconsin 
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin 
54301.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P. O. Box 1497, Madison, Wisconsin 53701-1497.
    NRC Project Director: John N. Hannon

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee 
Nuclear Power Plant, Kewaunee County, Wisconsin

    Date of amendment request: September 7, 1994
    Description of amendment request: The proposed amendment would 
revise Kewaunee Nuclear Power Plant (KNPP) Technical Specifications 
(TS) by adding two new sections, TS Section 3.0 and TS Section 4.0, 
with associated bases. TS Section 3.0 would establish the general 
requirements applicable to each of the Limiting Conditions for 
Operation (LCOs) within Section 3 of the KNPP TS. TS Section 4.0 would 
establish the general requirements applicable to Surveillance 
Requirements. The new requirements of TS 4.0.b would also affect TS 
Sections 4.5, 4.6, 4.7, and Tables TS 4.1-2 and 4.1-3. The proposed TS 
amendment incorporates guidance statements similar to Section 3.0/4.0 
of NUREG-0452, ``Standard Technical Specifications for Westinghouse 
Pressurized Water Reactors.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    The proposed changes were reviewed in accordance with the 
provision of 10 CFR 50.92 to show no significant hazards exist. The 
proposed changes will not:
    1) involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The likelihood that an accident will occur is neither increased 
or decreased by these TS changes. These TS changes will not impact 
the function or method of operation of plant equipment. Thus, there 
is not a significant increase in the probability of a previously 
analyzed accident due to these changes. No systems, equipment, or 
components are affected by the proposed changes. Thus, the 
consequences of the malfunction of equipment important to safety 
previously evaluated in the Updated Safety Analysis Report (USAR) 
are not increased by these changes.
    The proposed changes have no impact on accident initiators or 
plant equipment, and thus, do not affect the probabilities or 
consequences of an accident.
    These changes are consistent with the requirements established 
in the Westinghouse STS. Therefore, the proposed changes will not 
significantly increase 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 proposed TS changes would not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. The proposed changes do not involve changes to the 
physical plant or operations. Since these changes do not contribute 
to accident initiation, they do not produce a new accident scenario 
or produce a new type of equipment malfunction. Also, these changes 
do not alter any existing accident scenarios; they do not affect 
equipment or its operation, and thus, do not create the possibility 
of a new or different kind of accident.
    3) involve a significant reduction in the margin of safety.
    Operation of the facility in accordance with the proposed TS 
would not involve a significant reduction in a margin of safety. The 
proposed changes do not affect plant equipment or operation. Safety 
limits and limiting safety system settings are not affected by these 
proposed changes. These changes are consistent with the Westinghouse 
STS.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Wisconsin 
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin 
54301.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P. O. Box 1497, Madison, Wisconsin 53701-1497.
    NRC Project Director: John N. Hannon

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

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-
324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick 
County,North Carolina

    Date of amendments request: September 9, 1994 Brief description of 
amendments request: The amendments change the Technical Specifications 
to revise the frequency for verifying the position of the drywell-
suppression chamber vacuum breakers when the position indication is not 
operable from at least once every 72 hours to at least once every 14 
days.Date of publication of individual notice in Federal Register: 
September 16, 1994 (59 FR 47648)
    Expiration date of individual notice: October 3, 1994
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.

Gulf States Utilities Company, Cajun Electric Power Cooperative, 
and Entergy Operations, Inc., Docket No. 50-458, River Bend 
Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: September 8, 1994
    Brief description of amendment: The proposed amendment would modify 
Technical Specification 3.10.2, to permit the bypassing of the rod 
withdrawal limiter notch constraints while performing fuel power 
suppression testing. This modification to the technical specification 
will allow River Bend Station to search for and identify the location 
of leaking fuel bundles, during power operating conditions, so that 
appropriate actions can be taken to prevent further degradation.
    Date of publication of individual notice in Federal Register: 
September 16, 1994 (59 FR 47652)
    Expiration date of individual notice: October 17, 1994
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, Louisiana 70803

Gulf States Utilities Company, Cajun Electric Power Cooperative, 
and Entergy Operations, Inc., Docket No. 50-458, River Bend 
Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: September 12, 1994
    Brief description of amendment: The proposed amendment would revise 
the formula for calculating the average power range monitor (APRM) flow 
biased simulated thermal power-high reactor trip and flow biased 
neutron flux-upscale control rod block trip setpoints T-factor 
specified in Technical Specification (TS) 3/4.2.2. The proposed changes 
are necessary to support implementation of recommendations contained in 
NRC Generic Letter 94-02, ``Long-Term Solutions and Upgrade of Interim 
Operating Recommendations for Thermal-Hydraulic Instabilities in 
Boiling Water Reactors.''
    Date of publication of individual notice in Federal Register: 
September 21, 1994 (59 FR 48456)
    Expiration date of individual notice:  October 21, 1994
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, Louisiana 70803

Northeast Nuclear Energy Company, et al., Docket No. 50-423, 
Millstone Nuclear Power Station, Unit No. 3, New London County, 
Connecticut

    Date of amendment request: September 9, 1994
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications to modify surveillance requirements 
by increasing the acceptance criterion for the closure of the main 
steam isolation valves from 5 seconds to 10 seconds.
    Date of publication of individual notice in Federal Register: 
September 19, 1994 (59 FR 47960).
    Expiration date of individual notice: October 19, 1994
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, CT 06360.

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

    Date of application for amendments: September 8, 1994 (TS 94-14)
    Brief description of amendments: The amendment would separate the 
portion of the steam generator tubing from the end of the tube up to 
the start of the tube-to-tubesheet weld from the remainder of the tube 
for the purposes of sample selection and repair when defects are found 
in this section of a steam generator tube.
    Date of publication of individual notice in the Federal 
Register:September 19, 1994 (59 FR 47962)
    Expiration date of individual notice: October 19, 1994
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennesee 37402.

Notice Of Issuance Of Amendments To Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC 20555, and at the local public document 
rooms for the particular facilities involved.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 
50-529 and STN 50-530, Palo Verde Nuclear Generating Station, Unit 
Nos. 1, 2 and 3, Maricopa County, Arizona

    Date of application for amendments: August 5, 1993
    Brief description of amendments: The amendments change the phrase 
``Pressurizer Pressure - Wide Range'' to ``Reactor Coolant System 
Pressure - Wide Range'' in item 4 of TS Table 3.3-10 and item 4 of 
Table 4.3-7. These amendments will clarify the instrumentation required 
and eliminate potential confusion between the reactor coolant system 
pressure instruments and the pressurizer pressure instruments.
    Date of issuance: September 21, 1994
    Effective date: September 21, 1994
    Amendment Nos.: 81, 68, and 53
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 29, 1993 (58 
FR 50962) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 21, 1994. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318, 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of application for amendments: June 8, 1994
    Brief description of amendments: The amendments revise Technical 
Specification Section 4.7.1.2.c to extend the interval for three 
Auxiliary Feedwater surveillance requirements from 18 to 24 months.
    Date of issuance: September 26, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 197 and 174
    Facility Operating License No. DPR-53 and DPR-69: Amendment revised 
the Technical Specifications.
    Date of initial notice in Federal Register: August 17, 1994 (59 FR 
42334) The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated September 26, 1994.No 
significant hazards consideration comments received: No
    Local Public Document Room location:  Calvert County Library, 
Prince Frederick, Maryland 20678.

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318, 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of application for amendments: November 2, 1993, as 
supplemented on June 22, 1994
    Brief description of amendments: The amendments revise the 
Technical Specifications regarding surveillance requirements associated 
with the emergency diesel generators (EDGs) which include the 
following: 1) the surveillance interval is extended from 18 months to 
24 months which is the current refueling cycle; 2) removes the 
requirement to verify the EDGs speed; 3) exempts sequencer testing in 
Modes 5 and 6; 4) deletes the reference to the specific 2000 hour 
rating of the EDGs; and 5) allows the EDGs to be prelubricated prior to 
being started in accordance with the vendors recommendation.
    Date of issuance: September 27, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 198 and 175
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 8, 1993 (58 FR 
64599) The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated September 27, 1994.No 
significant hazards consideration comments received: No
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
PointNuclear Generating Unit No. 2, Westchester County, New York

    Date of application for amendment: April 29, 1994
    Brief description of amendment: The amendment revises surveillance 
intervals associated with initiation of auxiliary feedwater on steam 
generator water level (low-low) and on trip of the main feedwater 
pumps. These revisions are being made in accordance with the guidance 
provided by Generic Letter 91-04, ``Changes in Technical Specification 
Surveillance Intervals to Accommodate a 24-Month Fuel Cycle.''
    Date of issuance:  September 23, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 175
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 17, 1994 (59 FR 
42335) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 23, 1994.No 
significant hazards consideration comments received: No
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of application for amendment:  December 10, 1993, as 
supplemented by letter dated August 11, 1994.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) Section 5.3.A., ``Reactor Core,'' to allow the use 
of VANTAGE + fuel with ZIRLO cladding and of fuel with filler rods to 
permit fuel reconstitution. The amendment also revises the Basis for TS 
Section 2.1, ``Safety Limit: Reactor Core,'' to more accurately 
describe the basis of the departure from nucleate boiling correlations 
and how they are applied to ensure that the design criteria are met.
    Date of issuance:  September 29, 1994
    Effective date:  As of the date of issuance to be implemented 
within 30 days.
    Amendment No.: 176
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 2, 1994 (59 FR 
10003) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 29, 1994.No 
significant hazards consideration comments received: No
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

Dairyland Power Cooperative, Docket No. 50-409, La Crosse Boiling 
Water Reactor, La Crosse, Wisconsin

    Date of application for amendment: November 5, 1993 (Reference LAC-
13320) as supplemented August 3, 1994, (Reference LAC-13420).
    Brief description of amendment: This amendment modified the 
Technical Specifications (TS) incorporated in Possession-Only License 
No. DPR-45 in accordance with a revision of 10 CFR Part 20 (56 FR 
23360). In addition, there were minor clerical changes to correct 
oversights from previous amendments.
    Date of issuance: September 27, 1994.
    Effective date: This license amendment is effective as of the date 
of its issuance and must be fully implemented no later than 30 days 
from the date of issuance.
    Amendment No.: 68.Possession-Only License No. DPR-9: The amendment 
revised the TS.
    Date of initial notice in Federal Register: January 5, 1994 (59 FR 
618) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 27, 1994.No significant hazards 
consideration comments received: No.
    Local Public Document Room location: La Crosse Public Library, 800 
Main Street, La Crosse, Wisconsin 54601.

Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of application for amendments:  November 11, 1993, as 
supplemented February 23, April 12 and July 29, 1994.
    Brief description of amendments: The amendments reflect the 
consolidation of the Quality Verification Department with the Nuclear 
Generation Department that realigned the Nuclear Safety Review Board to 
report to the Senior Nuclear Officer, change a reference from Semi-
Annual to Annual, change an organizational unit term from ``group'' to 
``division,'' modify titles of positions designated to approve 
modifications and clarify the responsibilities of the Safety Assurance 
Manager.
    Date of issuance: September 23, 1994
    Effective date: September 23, 1994
    Amendment Nos.: 124 and 118
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 5, 1994 (59 FR 
618) The February 23, April 12 and July 29, 1994 letters provided 
clarifying information that did not change the scope of the November 
11, 1993, application and the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated September 23, 
1994.No significant hazards consideration comments received: No
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730

Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of application for amendments: November 11, 1993, and 
supplemented February 23, April 12 and July 29, 1994.
    Brief description of amendments: The amendments reflect the 
consolidation of the Quality Verification Department with the Nuclear 
Generation Department that realigned the Nuclear Safety Review Board to 
report to the Senior Nuclear Officer, change a reference from Semi-
Annual to Annual, change an organizational unit term from ``group'' to 
``division,'' modify titles of positions designated to approve 
modifications and clarify the responsibilities of the Safety Assurance 
Manager.
    Date of issuance: September 22, 1994
    Effective date: September 22, 1994
    Amendment Nos.: 148 and 130
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 19, 1994 (59 FR 
2865) The February 23, April 12 and July 29, 1994, letters provided 
clarifying information that did not change the scope of the November 
11, 1993, application and the initial proposed no significant hazards 
considerationdetermination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated September 22, 
1994. No significant hazards consideration comments received: No.
    Local Public Document Room location: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, 
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport, 
Pennsylvania

    Date of application for amendments:  June 2, 1994
    Brief description of amendments: These amendments revise the 
Appendix A TSs relating to reactor coolant leakage and leakage 
detection systems in an effort to bring TS sections 3/4.4.6.1 and 3/
4.4.6.2 closer to NRC's Improved Standard TSs. A new TS, Section 3/
4.5.5 for Unit 1 and 3/4.5.4 for Unit 2, is added to address Seal 
Injection Flow.
    Date of issuance:  September 22, 1994
    Effective date: September 22, 1994
    Amendment Nos.: 183 and 64
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 3, 1994 (59 FR 
39585) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 22, 1994No significant 
hazards consideration comments received: No.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam 
ElectricStation, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: February 9, 1993, as supplemented by 
letter dated July 22, 1994.
    Brief description of amendment: The amendment changed the Appendix 
A Technical Specifications by revising Specifications 3.0.4, 4.0.3, and 
4.0.4 in accordance with the intent of Generic Letter 87-09.
    Date of issuance:  September 20, 1994
    Effective date: September 20, 1994
    Amendment No.: 99
    Facility Operating License No. NPF-38. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 17, 1994 (59 FR 
42341) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 20, 1994.No 
significant hazards consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.

Georgia Power Company, Oglethorpe Power Corporation, Municipal 
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 
50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, 
Burke County, Georgia

    Date of application for amendments: April 28, 1994, and 
supplemented by letter dated July 29, 1994.
    Brief description of amendments: The proposed amendments would 
revise Technical Specification (TS) 3/4.8.1.1, ``AC Sources 
Operating,'' and the associated TS Bases for demonstrating the 
operability of the diesel generators (DGs), based upon the following 
NRC guidelines:A. Generic Letter (GL) 93-05, ``Line-Item Technical 
Specifications Improvements to Reduce Surveillance Requirements for 
Testing During Power Operation.'' B. Regulatory Guide (RG) 1.9, 
Revision 3, ``Selection, Design, Qualification, and Testing of 
Emergency Diesel Generator Units Used as Class 1E Onsite Electric Power 
Systems at Nuclear Power Plants,''
    Date of issuance: September 21, 1994
    Effective date: September 21, 1994
    Amendment Nos.: 75 and 54
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 3, 1994The 
Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated September 21, 1994.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Burke County Library, 412 
Fourth Street, Waynesboro, Georgia 30830

Georgia Power Company, Oglethorpe Power Corporation, Municipal 
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 
50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, 
Burke County, Georgia

    Date of application for amendments:  June 24, 1994.
    Brief description of amendments: The amendments revise the values 
of Z and S in Technical Specification 2.2-1 for the Pressurizer 
Pressure-Low and -High trip set-points (Table 2.2-1, Functional Units 9 
and 10) to allow the use of Tobar, Veritrak, or Rosemount pressure 
transmitters.
    Date of issuance: September 22, 1994
    Effective date: September 22, 1994
    Amendment Nos.: 76 and 55
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 22, 1994 (59 FR 
43143) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 22, 1994. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Burke County Library, 412 
Fourth Street, Waynesboro, Georgia 30830

Georgia Power Company, Oglethorpe Power Corporation, Municipal 
Electric Authority of Georgia, City of Dalton, Georgia, Docket No. 
50-321, Edwin I. Hatch Nuclear Plant, Unit 1, Appling County, 
Georgia

    Date of application for amendment: August 16, 1994, as supplemented 
September 20, 1994
    Brief description of amendments: The amendment makes a one-time 
change to Technical Specification (TS) 3.9.C for Hatch Unit 1 regarding 
the emergency diesel generator (DG) operability requirements during 
reactor shutdown conditions. Current TS 3.9.C requires that two DGs be 
operable during reactor shutdown when a core or containment cooling 
system is required to be operable. The amendment revises the current 
requirement such that only one emergency DG is required to be aligned 
to its associated core or containment cooling system during a specific 
time of the outage. During this time period the decay heat removal 
(DHR) system will be in service. The DHR system, which is completely 
independent of the existing shutdown cooling system, is powered by the 
Baxley substation and has its own DG as a backup power supply.
    Date of issuance: September 26, 1994
    Effective date: September 26, 1994
    Amendment Nos.: 194
    Facility Operating License Nos. DPR-57 and NPF-5. Amendment revised 
the Technical Specifications. The September 20, 1994, letter provided 
additional information that did not change the scope of the August 16, 
1994, application and the initial proposed no significant hazards 
consideration determination.
    Date of initial notice in Federal Register: August 26, 1994The 
Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated September 26, 1994.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513

GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station, Ocean County, New Jersey

    Date of application for amendment: April 19, 1994
    Brief description of amendment: The amendment updates and clarifies 
Technical Specification (TS) 3.4.B.1 to be consistent with TSs 1.39 and 
4.3.D. It addresses electromatic relief valve operability/bypassing 
during system pressure testing, including system leakage and 
hydrostatic test, with the reactor vessel solid, core not critical, and 
core reactivity limits satisfied.
    Date of issuance: September 27, 1994
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 170
    Facility Operating License No. DPR-16. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 25, 1994 (59 FR 
27056) The Commission's related evaluation of this amendment is 
contained in a Safety Evaluation dated September 27, 1994. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, New Jersey 
08753.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, 
Michigan

    Date of application for amendments: February 22, 1994
    Brief description of amendments: The amendments revise the 
Technical Specifications to reduce surveillance requirements for 
testing during power operation in the areas of control rod movement 
testing, radiation monitors, containment spray system, hydrogen 
recombiners, emergency diesel generators, special test exceptions - 
shutdown margin, and radioactive effluents - waste gas storage tanks.
    Date of issuance: September 28, 1994
    Effective date: September 28, 1994
    Amendment Nos.: 183 & 168
    Facility Operating License Nos. DPR-58 and DPR-74. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 30, 1994 (59 FR 
14890) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 28, 1994.No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.

Northeast Nuclear Energy Company, et al., Docket No. 50-423, 
Millstone Nuclear Power Station, Unit No. 3, New London County, 
Connecticut

    Date of application for amendment: April 18, 1994
    Brief description of amendment: The amendment revises the current 
surveillance frequency that verifies area temperature limits. The 
revised surveillance requirement will verify area temperature limits at 
least once per 7 days when the temperature monitor (datalogger) alarm 
is operable, and at least once per 12 hours when the datalogger alarm 
is inoperable.
    Date of issuance: September 22, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 95
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 3, 1994 (59 FR 
39593) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 22, 1994.No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Community-Technical College, Thames Valley Campus, 574 New London 
Turnpike, Norwich, Connecticut 06360.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, 
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis 
Obispo County, California

    Date of application for amendments:  August 29, 1994 (Reference LAR 
94-10)
    Brief description of amendments: The proposed amendments revise the 
combined Technical Specifications (TS) for the Diablo Canyon Power 
Plant Unit Nos. 1 and 2 to specify an alternate method of determining 
water and sediment content for new diesel fuel oil as specified in TS 
3/4.8.1.1, ``A.C. Sources - Operating.'' Specifically, TS 
4.8.1.1.3c.1(d) is revised to allow new fuel oil to be tested using a 
``clear and bright'' test or a quantitative test that verifies a water 
and sediment content less than or equal to 0.05 volume percent when the 
oil is tested in accordance with ASTM D1796-83.
    Date of issuance: September 23, 1994
    Effective date: September 23, 1994
    Amendment Nos.: 95 and 94
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.Public comments requested as to 
proposed no significant hazards consideration: Yes (59 FR 46453, dated 
September 8, 1994). The notice provided an opportunity to submit 
comments on the Commission's proposed no significant hazard 
consideration determination. No comments have been received. The notice 
also provided for an opportunity to request a hearing by October 7, 
1994, but stated that, if the Commission makes a final no significant 
hazards consideration determination, any such hearing would take place 
after issuance of the amendments.
    The Commission's related evaluation of the amendments, finding of 
exigent circumstances, and final determination of no significant 
hazards consideration is contained in a Safety Evaluation dated 
September 23, 1994.
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P. O. Box 7442, San Francisco, California 94120
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407

Southern California Edison Company, et al., Docket Nos. 50-361 and 
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, 
San Diego County, California

    Date of application for amendments:  October 29, 1992
    Brief description of amendments: These amendments revise the 
Technical Specification by adding an alternate method of ensuring that 
power to the safety injection tank vent valves is removed. The existing 
method verifies that the fuses are removed. The alternate method 
verifies that the disconnect switches are in the open position.
    Date of issuance: September 27, 1994
    Effective date: As of the date of its issuance.
    Amendment Nos.: 112 and 101
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 17, 1993 (58 
FR 8783) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 27, 1994.No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, 
Alabama

    Date of application for amendments: April 4, 1994 (TS 322)
    Brief description of amendment: The amendments eliminate the 
requirements in the Technical Specifications (TS) for automatic 
actuation of the following functions upon Main Steamline Radiation 
Monitor (MSRM) detection of a high radiation condition in the main 
steamlines:(1) reactor scram (2) main steam isolation valve closure(3) 
main steam line drain valve closure(4) reactor recirculation sample 
line valve closure(5) main condenser mechanical vacuum pump isolation 
and trip
    Date of issuance: September 27, 1994
    Effective date: September 27, 1994
    Amendment Nos.: 212, 227 and 185
    Facility Operating License Nos. DPR-33, DPR-52 and DPR-68: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: June 8, 1994 (59 FR 
29636) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 27, 1994.No 
significant hazards consideration comments received: None
    Local Public Document Room location: Athens Public Library, South 
Street, Athens, Alabama 35611

The Cleveland Electric Illuminating Company, Centerior Service 
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania 
Power Company, Toledo Edison Company, Docket No. 50-440, Perry 
Nuclear Power Plant, Unit No. 1, Lake County, Ohio

    Date of application for amendment: March 19, 1992
    Brief description of amendment: This amendment revised Technical 
Specifications to incorporate clarifications and corrections. These 
changes were administrative and not safety significant.
    Date of issuance: September 21, 1994
    Effective date: date of issuance, to be implemented within 90 days
    Amendment No. 66
    Facility Operating License No. NPF-58. This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 8, 1992 (57 FR 
30260) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 21, 1994.No 
significant hazards consideration comments received: No
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081.

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, 
Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks, 
Manitowoc County, Wisconsin

    Date of application for amendments: May 26, 1994 as supplemented 
July 11, 1994, and August 1, 1994.
    Brief description of amendments: Point Beach Nuclear Plant is 
installing two additional emergency diesel generators and reconfiguring 
portions of the 4160-Volt emergency electrical power system. The 
amendment revised the Point Beach Nuclear Plant Technical 
Specifications (TS) to establish the requirements for the electrical 
systems at Point Beach such that the TS will provide the appropriate 
guidance for all interim configurations and the final configuration. 
The majority of changes were incorporated in TS Section 15.3.7, 
``Auxiliary Electrical Systems.'' Other Sections modified were 15.3.0, 
``General Considerations,'' 15.3.14, ``Fire Protection System,'' and 
15.4.6, ``Emergency Power System Periodic Tests.''
    Date of issuance:  September 23, 1994
    Effective date: immediately, to be implemented within 45 days
    Amendment Nos.: 152 and 156
    Facility Operating License Nos. DPR-24 and DPR-27. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 20, 1994 (59 FR 
37092) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 23, 1994.The July 11, 
1994, and August 1, 1994, submittals provided additional supplemental 
information that did not change the initial proposed no significant 
hazards consideration determination.No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241.

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, 
Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks, 
Manitowoc County, Wisconsin

    Date of application for amendments: August 9, 1994, as supplemented 
on August 19, 1994.
    Brief description of amendments: These amendments revised the 
Technical Specifications (TS) 5.3.4, ``Steam and Power Conversion 
Systems,'' and TS 15.3.7, ``Auxiliary Electrical Systems,'' to increase 
the allowed outage times for one motor driven auxiliary feedwater pump 
and for the standby emergency power for the Unit 1, Train B4160 Volt 
safeguards bus (A06) from 7 to 12 days. The amendments also modified TS 
15.3.3, ``Emergency Core Cooling System, Auxiliary Cooling Systems, Air 
Recirculation Fan Coolers, and Contained Spray,'' to provide the 
clarification that the service water pump (P-32E) operating with power 
supplied by the Alternative Shutdown System is operable from offsite 
power. The changes are one-time extensions of specific allowed outage 
times.
    Date of issuance: September 23, 1994
    Effective date: immediately, to be implemented within 45 days
    Amendment Nos.: 153 and 157
    Facility Operating License Nos. DPR-24 and DPR-27. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 19, 1994 (59 FR 
42870) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 23, 1994.The August 
19, 1994, submittal provided additional supplemental information that 
did not change the initial proposed no significant hazards 
consideration determination.No significant hazards consideration 
comments received: No.
    Local Public Document Room location: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241.

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, 
Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks, 
Manitowoc County, Wisconsin

    Date of application for amendments: July 18, 1994
    Brief description of amendments: The amendments changed Technical 
Specification 15.3.7, ``Auxiliary Electrical System'' to include the 
allowed outage time for one of the four connected station battery 
chargers and subsequent shutdown requirements. The amendments also 
revised the basis for Section 15.3.7 to support the above changes.
    Date of issuance: September 29, 1994
    Effective date: immediately, to be implemented within 45 days
    Amendment Nos.: 154 and 158
    Facility Operating License Nos. DPR-24 and DPR-27. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 17, 1994 (59 FR 
42348) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 29, 1994.No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241.

Notice Of Issuance Of Amendments To Facility Operating Licenses And 
Final Determination Of No Significant Hazards Consideration And 
Opportunity For A Hearing (Exigent Public Announcement Or Emergency 
Circumstances)

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC 20555, 
and at the local public document room for the particular facility 
involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By 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. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington, DC 20555, by the 
above date. Where petitions are filed during the last 10 days of the 
notice period, it is requested that the petitioner promptly so inform 
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
operator should be given Datagram Identification Number N1023 and the 
following message addressed to (Project Director): petitioner's name 
and telephone number, date petition was mailed, plant name, and 
publication date and page number of this Federal Register notice. A 
copy of the petition should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and 
to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

Northeast Nuclear Energy Company, et al., Docket No. 50-423, 
MillstoneNuclear Power Station, Unit No. 3, New London County, 
Connecticut

    Date of application for amendment: September 17, 1994
    Brief description of amendment: The amendment revises the Technical 
Specifications (TS) Surveillance Requirements 4.3.2.2, 4.6.3.1, 
4.7.1.5.2, and 4.7.1.2.1.b by noting that surveillance requirement 
4.0.4 is not aplicable. The amendment allows the plant to enter Modes 4 
and 3, as necessary, to perform the required operability tests for the 
Main Steam Isolation Valves, the engineered safety feature actuation 
system and the turbine-driven Auxiliary Feedwater pump.
    Date of issuance: September 29, 1994
    Effective date: September 29, 1994
    Amendment No.: 96
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.Public comments requested as to proposed no 
significant hazards consideration: No. The Commission's related 
evaluation of the amendment, finding of emergency circumstances, and 
final determination of no significant hazards consideration are 
contained in a Safety Evaluation dated September 29, 1994.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, CT 06360.
    Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT 
06141-0270.
    NRC Project Director: John F. Stolz

Washington Public Power Supply System, Docket No. 50-397, Nuclear 
Project No. 2, Benton County, Washington

    Date of application for amendment:  September 18, 1994
    Brief description of amendment: The amendment modifies the 
Technical Specifications (TS) to add a note to TS Table 3.6.3-1, 
``Primary Containment Isolation Valves,'' to allow operation of the 
facility until the next plant shutdown, but not later than May 15, 
1995, without meeting the single-failure criterion for the logic 
circuit for containment isolation valves in the hydraulic lines 
supplying motive force for the reactor recirculation system (RRC) flow 
control valves.
    Date of issuance: September 29, 1994
    Effective date: September 29, 1994
    Amendment No.: 132
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications. Public comments on proposed no significant 
hazards consideration comments received: No. The Commission's related 
evaluation of the amendment, finding of emergency circumstances, and 
final determination of no significant hazards consideration are 
contained in a Safety Evaluation dated September 29, 1994.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352
    Attorney for licensee: M.H. Philips, Jr., Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005-3502.
    NRC Project Director: Theodore R. Quay
    Dated at Rockville, Maryland, this 4th day of October 1994.
    For the Nuclear Regulatory Commission
Jack W. Roe,
Director, Division of Reactor Projects - III/IV, Office of Nuclear 
Reactor Regulation
[Doc. 94-25024 Filed 10-11-94; 845 am]
BILLING CODE 7590-01-F