[Federal Register Volume 60, Number 70 (Wednesday, April 12, 1995)]
[Notices]
[Pages 18621-18640]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X95-50412]



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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 March 17, 1995, through March 31, 1995. The 
last biweekly notice was published on March 29, 1995 (60 FR 16181).

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 [[Page 18622]] 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. 
The filing of requests for a hearing and petitions for leave to 
intervene is discussed below.
    By May 12, 1995, the licensee may file a request for a hearing with 
respect to issuance of the amendment to the subject facility operating 
license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
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, 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 [[Page 18623]] 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, and at the local public document room for 
the particular facility involved.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, IllinoisDocket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, 
Will County, Illinois

    Date of amendment request: March 23, 1994, as supplemented on July 
26, 1994, February 15, 1995, and February 28, 1995.
    Description of amendment request: In the submittals of March 23 and 
July 26, 1994, the licensee requested revisions to the plants' 
technical specifications (TSs) to permit the use of a slightly positive 
reactor core moderator temperature coefficient (MTC). The February 15, 
1995, submittal requested approval to expand the operating limits 
report (OLR) to include a cycle specific MTC value and requested 
approval to maintain the MTC value within the limits specified in the 
OLR. The maximum upper MTC limit would be specified in the TSs. The 
February 28, 1995, submittal provided a revised Significant Hazards 
Consideration. This supplements the information that was published in 
the Federal Register on August 31, 1994 (59 FR 45037).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    An analysis program was pursued by Commonwealth Edison to 
justify a positive MTC, reduced reactor coolant system thermal 
design flow, and increased steam generator tube plugging levels. 
This analysis identified a need for corresponding increases in the 
boron concentration levels in the refueling water storage tank 
(RWST) and safety injection accumulators to assure subcriticality 
requirements are met following a postulated loss-of-coolant accident 
(LOCA). The increases in boron concentration are based on the 
maximum upper limit of the MTC. The corresponding Technical 
Specification changes required as a result of this analysis program 
were previously approved by the NRC, including the increases in 
boron concentration limits, with the exception of the positive MTC 
change. The safety analyses necessary to support this program are 
documented in WCAP-13964. The results were reviewed by Commonwealth 
Edison and found to be acceptable. All Departure from Nucleate 
Boiling Ratio (DNBR) design limits were determined such that there 
was a 95 percent probability at a 95 percent confidence level that 
DNB would not occur on the most limiting fuel rod for any Condition 
I or Condition II event. The present Technical Specification limit 
for Nuclear Enthalpy Rise Hot Channel Factor, ... , of less than 
1.65 ensures that the DNB design basis stated above would be met, 
thus fuel integrity will not be challenged.
    The accidents which are sensitive to MTC were analyzed as part 
of the overall program and the results were found to be acceptable. 
The safety functions of the evaluated systems and components remain 
unchanged. The analysis performed using the increased MTC value does 
not affect the integrity of the safety related systems and 
components such that their function to control radiological 
consequences is affected and all fission barriers will remain 
intact. The effects on offsite doses have been considered. The 
incorporation of a positive MTC, in conjunction with the previously 
approved reduction in reactor coolant system thermal design flow 
rate and increase in steam generator tube plugging levels, will 
result in a small increase in offsite doses; however, the total 
doses remain a small fraction of the 10 CFR 100 limits. As such, the 
accident analysis acceptance criteria continue to be satisfied.
    On a cycle-by-cycle basis, a deterministic evaluation of the 
impact on ATWS risk will be performed. An Unfavorable Exposure Time 
(UET) will be calculated, where UET is defined as the amount of time 
during the operating cycle for which the reactivity feedback is not 
sufficient to prevent Reactor Coolant System (RCS) pressure from 
exceeding 3200 psig for a given plant configuration. The UET 
methodology is consistent with the Westinghouse Owner's Group 
methodology presented in WCAP 11992, ``ATWS Rule Administration 
Process'' and WCAP 11993, ``Assessment of Compliance with ATWS Rule 
Basis for Westinghouse PWRs''. Corrective actions will be taken, as 
necessary, to assure a UET of less than 5 percent of cycle length.
    The relocation of the cycle-specific core operating limits for 
the MTC from the Technical Specifications has no influence or impact 
on the probability or consequences of any accident previously 
evaluated. Byron and Braidwood Stations will continue to operate 
within the cycle-specific MTC limits contained in the OLR. The 
proposed amendment will require exactly the same action to be taken 
when the OLR limits are exceeded as are required by the current 
Technical Specification. Any change to the MTC values in the OLR 
will be performed based on NRC-approved methodology as delineated in 
Section 6.9.1.9 of the Technical Specifications. Each accident 
analysis addressed in the Updated Final Safety Analysis Report 
(UFSAR) will be examined with respect to changes in cycle dependent 
parameters, which are obtained from application of NRC-approved 
reload design methodologies, to ensure that the transient evaluation 
of new reloads are bounded by previously accepted analysis. This 
examination, which will be performed under the requirements of 10 
CFR 50.59, ensures that future reloads will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    Therefore, implementation of a positive MTC will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different type of accident from any accident previously 
evaluated.
    The methodology and manner of plant operation as a result of the 
proposed changes is unaffected. Implementation of a positive MTC 
does not impact the safe operation of the reactor provided that the 
Limiting Conditions for Operation (LCOs) and the associated action 
requirements are satisfied. The assumptions do not create any new 
failure modes that could adversely impact safety related equipment. 
The reload safety limits and LCOs in the plant Technical 
Specifications will be evaluated and satisfied for each future 
reload core design via the 10 CFR 50.59 process. All DNBR limits 
have been satisfied. Currently installed equipment will not be 
operated in a manner different than previously designed. No new 
credible limiting single failure has been created. No new or 
different accidents or failure modes have been identified for any 
systems or components important to safety.
    The relocation of the cycle specific MTC values to the OLR will 
not create the possibility of a new or different type of accident. 
No safety related equipment or safety function will be altered as a 
result of this proposed change. The cycle specific values are 
calculated using NRC-approved methods and submitted to the NRC to 
allow the Staff to continue to trend these limits. The Technical 
Specifications will continue to require operation within the 
analyzed core operating limits and appropriate actions will be 
taken, when, or if, the limits are exceeded. [[Page 18624]] 
    Therefore, there is not a potential for creating the possibility 
of a new or different type of accident from any accident previously 
evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The performance and integrity of the evaluated safety related 
systems and components are not affected by the proposed change to 
the MTC. The radiological consequences of all previously analyzed 
accidents remain within acceptable limits. The proposed change to 
the MTC will have no effect on the availability, operability, or 
performance of the evaluated safety related systems or components. 
The incorporation of a positive MTC, in conjunction with the 
previously approved reduction in reactor coolant system thermal 
design flow rate and increase in steam generator tube plugging 
levels, will result in a small increase in offsite doses; however, 
the total doses remain a small fraction of the 10CFR100 limits. The 
methodology, discussed in Attachment E, describes the determination 
and use of the UET values in the calculation of the Primary Pressure 
Relief node for the ATWS event tree to determine an overall ATWS 
risk value. The methodology will be used by ComEd to ensure that a 
core designed with a positive MTC will not result in an unacceptable 
risk to core damage frequency due to an ATWS event. The margin of 
safety associated with the licensing basis safety analysis is not 
significantly reduced by the proposed changes. All acceptance 
criteria for the specific UFSAR Chapter 15 safety analyses (non-LOCA 
and LOCA) have been satisfactorily evaluated and verified using NRC 
approved methodologies.
    The margin of safety is not affected by the relocation of the 
cycle specific MTC limits from the Technical Specifications. The 
proposed amendment continues to require operation within the core 
limits as determined by the NRC-approved reload design and safety 
analysis methodologies. Appropriate actions will be taken, when, or 
if, limits are exceeded.
    The development of the MTC limits for future reloads will 
continue to conform to those methods described in the NRC-approved 
documentation. In addition, each future reload will involve 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 the margin of 
safety as defined in the basis for any Technical Specification.
    Therefore, there is no significant reduction in the margin of 
safety as defined in the bases of any Technical Specification.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: For Byron, the Byron Public 
Library, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for 
Braidwood, the Wilmington Township Public Library, 201 S. Kankakee 
Street, Wilmington, Illinois 60481.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690
    NRC Project Director: Robert A. Capra

Consumers Power Company, Docket No. 50-255, Palisades Plant, Van 
Buren County, Michigan

    Date of amendment request: October 20, 1992
    Description of amendment request: The proposed amendment would 
comply with the requirements of Amendment 135 to the Palisades 
Operating License, dated February 11, 1991, which included a change to 
Technical Specification 5.3.1a, Primary Coolant System. The safety 
evaluation for Amendment 135 included a requirement that changes to 
Section 4.2 of the Palisades Final Safety Analysis Report (FSAR) be 
made through a formal amendment process. The proposed FSAR change is a 
result of the steam generator replacement project and includes the 
following: (1) deletion of a design load since this was not treated as 
a necessary design condition in the new steam generators; (2) a change 
in the feedwater temperature from 70 deg.F to 40 deg.F, since this 
assumption was changed in the analysis for the replacement steam 
generators; and (3) editorial 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 following summary supports the finding that the proposed 
change would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The probability of an accident previously evaluated in the FSAR 
will not be increased by deleting the design load change of 15% per 
minute or decreasing the minimum feedwater temperature from 70 deg.F 
to 40 deg.F. There is no design requirement that the plant be 
capable of 15% per minute load changes. No accident has as an 
initial condition a 15% per minute load change taking place, and 
since this FSAR change is the result of the replacement steam 
generators design, no accident probabilities are increased. The 
40 deg.F feedwater temperature affects the steam generators, but 
nothing else is affected in the primary coolant system (PCS). The 
replacement steam generators have been shown by the design analysis 
report to be able to withstand the same number of cycles of the 
addition of 40 deg.F water as the old steam generators could with 
70 deg.F water.
    The consequences of an accident previously evaluated in the FSAR 
are not increased by either of these two changes. Deleting the 
design load rate of 15% per minute deals with normal plant operation 
and would not affect the course of a Chapter 14 event since none of 
the Chapter 14 events involve power level changes with respect to 
the steam generators. Also, reducing the maximum design load change 
rate is a conservative change.
    Lowering the feedwater temperature could increase the 
consequences of the main steam line break (MSLB) accident by 
increasing the likelihood of a return to power event caused by 
increased core cooling; however, the current FSAR analysis in 
Section 14.14 used 32 deg.F as the auxiliary feedwater temperature 
and thus bounds [the] 40 deg.F [temperature].
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The possibility of a new or different type of accident is not 
created by these FSAR changes. By deleting the 15% per minute load 
change rate from the FSAR, the operation of the plant is unaffected 
because the 5% per minute limit on load rate change is more 
limiting. There is no license requirement to be able to change power 
at 15% per minute except as described in the proposed FSAR deletion. 
Furthermore, FSAR Section 4.3.7.2 states that the pressurizer 
heaters cannot be uncovered by the outward surge of water following 
load increases; a 10% step increase and 15% ramp increase. FSAR 
Section 1.2.4.9.a states that the nuclear steam supply system (NSSS) 
is capable of a ramp change from 15% to 100% power at 5% per minute, 
and at a greater rate over smaller load changes up to a step change 
of 10%.
    Another consideration is that the analysis for the original 
steam generators was not as detailed or exact as the analysis for 
the replacement steam generators. The thermal analysis section of 
the original steam generator design analysis report states for the 
three power change cases, 5% per minute, 15% per minute and a 10% 
step change, that ''... the transient thermal effects of the power 
changes are small and [negligible]. The situations of significance 
are due to cycling between steady state conditions at different 
power levels.'' Thus, the rate of change was not a consideration in 
the original design analysis. The replacement steam generator 
analysis calculated the transient temperature changes with respect 
to time, so the rate of change was considered. Therefore, the 
replacement steam generator analysis is more accurate, but does not 
consider a 15% per minute rate change. The original steam generators 
were not designed for 15% per minute power changes but could 
withstand power increases from 50% to 100% [a total of] 15,000 times 
without considering the rate of power change.
    Reducing the analyzed feedwater temperature from 70 deg.F to 
40 deg.F does not change the possibility of whether another type of 
accident or malfunction can occur since the steam generator is 
analyzed for this.
    3. Involve a significant reduction in a margin of safety.
    The margin of safety as defined by plant licensing basis is not 
reduced due to the replacement steam generators not being analyzed 
for a 15% per minute power ramp [[Page 18625]] because the 15% per 
minute ramp rate was not a licensing basis of the plant design. The 
original plant Safety Evaluation Report does not mention the design 
power ramp rates. The basis for Technical Specification 3.1.2 states 
that all components are designed to withstand the effects of cyclic 
loads due to primary coolant system temperature and pressure changes 
induced by load changes, trips, and start-ups and shutdowns. FSAR 
Section 4.2.2 is referenced. The change of eliminating the analyzed 
ability to make 15% per minute power changes does not reduce the 
margin of safety because:
    a. the plant is not operated in a manner wherein 15% per minute 
power increases are made. Rapid power decreases during emergency 
conditions are not covered by this analysis since they are not 
controlled to 15% per minute but should be considered analyzed by 
the 500 trips or 10% step change analysis and,
    b. the original steam generator did not use the ramp rate in the 
analysis and,
    c. a 15% per minute power change from 50% to 100% power is a 
fairly benign change for the steam generator with respect to 
pressure and temperature changes as compared to heatups and 
cooldowns because the total changes are small.
    The only requirement from the NRC with respect to the number and 
type of loads is contained in Section II of the NRC Standard Review 
Plan (SRP) 3.9.1 which states ''...The section of the applicant's 
SAR which pertains to transients will be acceptable if the transient 
conditions selected for equipment fatigue evaluation are based upon 
a conservative estimate of the magnitude and frequency of the 
temperature and pressure conditions resulting from those 
transients.'' ''... Transients and resulting loads and load 
combinations with appropriate specified design and service limits 
must provide a complete basis for design of the reactor coolant 
pressure boundary for all conditions and events expected over the 
service lifetime of the plant.''
    In the intervening years between design of the original steam 
generators and the replacement steam generators, Combustion 
Engineering (ABB-CE) decided that a 15% per minute power ramp rate 
was beyond what was necessary and expected to occur. This position 
was acceptable to the NRC since ABB-CE letter CPC-90-170, dated 
October 24, 1990, states that the replacement steam generators are 
identical in design to the Palo Verde (Arizona Public Service) steam 
generators. (The ABB-CE letter was concerned with the stress 
analysis for steam line breaks, therefore, the reference to being 
identical was with respect to that stress analysis.)
    The change in feedwater temperature from 70 deg.F to 40 deg.F 
maintains the margin of safety because the replacement steam 
generators have been shown by the design analysis report to be able 
to withstand the same number of cycles of the addition of 40 deg.F 
water as the old steam generators could 70 deg.F water.
    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: Van Wylen Library, Hope 
College, Holland, Michigan 49423.
    Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201
    NRC Project Director: Cynthia A. Carpenter, Acting

Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County, 
Michigan

    Date of amendment request: September 13, 1993
    Description of amendment request: The proposed amendment would 
relocate audit frequencies of Section 6.5.2.8 of the Technical 
Specifications to the Quality Assurance Program in Section 17.2 of the 
Updated Final Safety Analysis Report.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed change to relocate the audit program frequency 
requirements to the Quality Assurance Program does not:
    (1) involve a significant increase in the probability or 
consequences of an accident previously evaluated,
    This change is administrative in nature and does not impact the 
operation of the plant or the plant's response to an accident. 
Because it will allow more flexibility in assigning resources to 
assess weak or declining performance areas, the plant safety 
performance will be improved.
    (2) create the possibility of a new or different kind of 
accident from any accident previously evaluated,
    This change is administrative in nature and does not affect the 
operation or design of the plant; therefore, there is no change in 
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.
    This change is administrative in nature and does not affect the 
operation of the plant; therefore, there is no change in the margin 
of safety. Relocating the audit program frequency requirements to 
the Quality Assurance program will allow a more dynamic and 
responsive audit program. Audits will be able to be scheduled more 
effectively based on performance and the status of related 
activities. This should result in a more effective audit program 
that will contribute to an improvement in 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: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161
    Attorney for licensee: John Flynn, Esq., Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226
    NRC Project Director: Cynthia A. Carpenter, Acting

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, 
ArkansasNuclear One, Unit Nos. 1 and 2 (ANO-1&2), Pope County, 
Arkansas

    Date of amendment request: August 30, 1994, with supplement dated 
January 19, 1995.
    Description of amendment request: The proposed amendment changes 
requirements related to the site parimeter security 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, excerpts of this analysis are presented below:
    Criterion 1 - Does Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated
    The accident mitigation features of the plant are not affected 
by the proposed change. This change provides an equivalent level of 
protection as required by 10CFR73.55(c)(4), does not significantly 
decrease the effectiveness of the security program, and is adequate 
for preventing an unacceptable risk to public health and safety. 
Ample protection against a design basis security threat continues to 
be provided. Therefore, the probability or consequences of an 
accident previously evaluated is not significantly increased.
    Criterion 2 - Does Not Create the Possibility of a New or 
Different Kind of Accident from Any Previously Evaluated
    This change clarifies the existing configuration of the 
protected area barrier at the ANO intake structure. New systems, 
modes of equipment operation, failure modes, or other plant 
perturbations are not introduced by this change. Therefore, the 
possibility of a new or different kind of accident from amy 
previously evaluated is not created.
    Criterion 3 - Does Not Involve a Significant Reduction in the 
Margin of Safety
    This change clarifies the existing configuration of the 
protected area barrier at the ANO intake structure. The proposed 
change does not alter a safety limit, a limiting condition of 
operation, or a surveillance requirement on equipment to operate the 
plant. Adequate physical protection of the plant is maintained. 
Therefore, the margin of safety is not significantly 
reduced. [[Page 18626]] 
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center 
(DAEC), Linn County, Iowa

    Date of amendment request: March 1, 1995
    Description of amendment request: The proposed License Amendment 
would revise Technical Specification (TS) Sections 4.5 and 4.8 of the 
DAEC TS to reflect the changes to pump and valve testing criteria. The 
proposed amendment changes the testing frequency for certain pumps and 
valves in the Low Pressure Coolant Injection subsystem; Core Spray 
subsystems; and the Residual Heat Removal Service Water, High Pressure 
Coolant Injection, Emergency Service Water, and River Water Supply 
systems. The frequency would change from testing every three months to 
that specified by DAEC ASME Section XI Inservice Testing (IST) program.
    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 affected pumps and valves in Sections 4.5 and 4.8 will 
continue to be tested in accordance with ASME Section XI OM-6 and 
OM-10. The affected pumps and valves will continue to function as 
before and this change will not result in a decrease in their 
availability to mitigate the consequences of certain accidents and 
transients. The proposed amendment will not affect the consequences 
of these accidents and transients. Therefore, the
    proposed amendment does not involve a change in the probability 
or consequences of an accident previously evaluated.
    (2) The proposed license amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated. The safety functions of the affected pumps and 
valves will remain unchanged. This amendment will result in no 
physical changes to the affected pumps, valves or systems. 
Consequently, the proposed license amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    (3) The proposed amendment will not reduce the margin of safety. 
The actual operation of the affected pumps and valves will remain 
unchanged. Testing in accordance with ASME Section XI OM-6 and OM-10 
will continue to provide assurance that degradation in tested 
components will be detected and addressed.
    The NRC staff has reviewed the licensee's analysis and, based on 
thisreview, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, S.E., Cedar Rapids, Iowa 52401
    Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan, Lewis 
& Bockius, 1800 M Street, N.W., Washington, DC 20036-5869NRC Acting 
Project Director: Gail H. Marcus

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: January 24, 1995, as supplemented March 
22, and March 29, 1995.
    Description of amendment request: The amendment request would 
revise the Technical Specification Section 3.2.3.1.a and Table 2.2-1 to 
decrease the acceptance criterion for measured reactor coolant system 
(RCS) flow rate from 387,480 gallons per minute (gpm) to 371,920 gpm.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration (SHC), which is presented below:
    ...The proposed changes do not involve an SHC because the 
changes would not:
    1. Involve a Significant Increase in the Probability or 
Consequence of an Accident Previously Evaluated.
    An evaluation of the 4% decrease in the RCS total flow rate 
limit has shown that the change does not significantly impact the 
design basis analyses. Therefore, the change will not increase the 
consequences of an accident previously evaluated.
    There are no actual plant changes that will result from this 
technical specification change. Instead, the technical specification 
requirement for minimum total RCS flow rate is being changed to 
provide operational benefit without compromising safety. Since there 
are no plant changes, there is no effect on the probability of 
occurrence of previously evaluated accidents.
    The change will have a negligible impact on the small break loss 
of coolant accident (LOCA) and large break LOCA analyses. The PCT 
[peak cladding temperature] acceptance criteria will continue to be 
met with the assumption of a 4% reduction in RCS flow rate.
    For the steam generator tube rupture event, both the FSAR [Final 
Safety Analysis Report] offsite dose analysis and the margin of 
steam generator (SG) overfill were evaluated. It was determined that 
the 4% reduction in RCS flow rate will not adversely affect the 
offsite doses or the margin to SG overfill and, therefore, the FSAR 
conclusions remain unchanged.
    In the evaluation of non-LOCA transients, the DNB [departure 
from nucleate boiling] is the most affected parameter due to a 
change in flow rate. It was concluded that the 4% reduction in RCS 
flow was acceptable and there was margin to the DNB limit.
    It is concluded that there is sufficient margin to the system 
pressure, PCT and DNB limits to offset the effect of the 4% flow 
rate decrease and the calculated radiological releases associated 
with the analysis are not affected. Therefore, there is no effect on 
the consequences of previously evaluated accidents.
    2. Create the Possibility of a New or Different Kind of Accident 
from any Previously Evaluated.
    The low loop flow trip setpoint specified in Technical 
Specification Table 2.2-1 is set as a fraction of total flow. The 
flow fraction is not being changed and no hardware changes are 
required due to the reduction in minimum flow. Also, the reduction 
in minimum flow will not change the operation of any plant equipment 
and it does not modify plant operation.
    Therefore, the reduction in minimum flow does not introduce any 
new failure modes or malfunctions and it does not create the 
potential for a new unanalyzed accident.
    3. Involve a Significant Reduction in the Margin of Safety.
    The proposed 4% decrease in the technical specification limit 
for total RCS flow rate will not adversely affect the results of the 
FSAR accident analysis, and it is concluded that this change is 
safe. The change does not adversely affect any equipment credited in 
the safety analysis, and it does not affect the probability of 
occurrence of any plant accident. Also, the change has a negligible 
impact on the PCT, and it does not increase the offsite doses or 
decrease the DNB below its acceptance limit.
    Therefore, the change does not have any significant impact on 
the protective boundaries, and there is no reduction in the margin 
of safety as specified in the technical specifications.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 
[[Page 18627]] 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: Phillip F. McKee

Omaha Public Power District, Docket No. 50-285, Fort Calhoun 
Station,Unit No. 1, Washington County, Nebraska

    Date of amendment request: March 1, 1995
    Description of amendment request: The proposed amendment to the 
technical specifications (TS) would make administrative changes to TS 
2.5, 2.8, 2.11, 3.2, and 3.10 and, in accordance with Generic Letter 
(GL) 93-07, ``Modification of the Technical Specification 
Administrative Control Requirements for Emergency and Security Plans,'' 
to TS 5.5 and 5.8.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) The proposed changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed revisions to Technical Specifications (TS) 5.5 and 
5.8 are administrative in nature and follow the guidance of Generic 
Letter (GL) 93-07. The review and audit functions of the site 
security and emergency plans and procedures will be retained in a 
manner that fully satisfies regulatory requirements. Therefore, the 
proposed revisions do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed revision to TS 2.5 will still require backup water 
for the emergency feedwater storage tank to be available. However, 
several other available sources of water are preferred over river 
water, such as, the water plant demineralized water system and the 
outside condensate storage tank. Therefore, the proposed revision 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed deletion of TS 2.8(8) pertaining to fuel handling 
cranes, deletion of TS 2.11 pertaining to overhead cranes in the 
Containment and Auxiliary Buildings, and deletion of statements in 
the bases of TS 2.8 pertaining to crane interlocks does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated. Specifications 2.8(8), 2.11 and the 
deleted statements in the bases of Specification 2.8 need not be 
retained in the TS based upon Criteria 1 through 4 of the ``Final 
Policy Statement on Technical Specifications Improvements for 
Nuclear Power Reactors,'' dated July 22, 1993 (58 FR 39132).
    Controls and limitations for the operation and testing of these 
cranes and interlocks will be incorporated into the Updated Safety 
Analysis Report (USAR). The requirements of TS 2.8(8) and 
restrictions of TS 2.11 are currently contained in Station 
procedures to ensure that the handling of fuel assemblies, control 
element assemblies (CEAs) and heavy loads is accomplished safely and 
effectively. These revisions make the FCS Technical Specifications 
more similar to Standard Technical Specifications (STS), which do 
not contain requirements or restrictions concerning the operation of 
fuel handling cranes or overhead cranes.
    The revision proposed for TS 3.2, Table 3-5, Item 1 will make 
its surveillance frequency identical to the frequency specified in 
STS 3.1.5.7. The proposed frequency will require testing CEA drop 
times prior to reactor criticality after each removal of the reactor 
vessel closure head, which is the most appropriate time to perform 
the surveillance. The proposed frequency will ensure that the CEAs 
drop into the core within the time specified in the safety analysis 
and, therefore, does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed deletion of TS 3.2, Table 3-5, Item 5, which 
currently requires testing refueling system interlocks prior to the 
refueling outage does not involve a significant increase in the 
probability or consequences of an accident previously evaluated. 
Table 3-5, Item 5, does not need to be retained in the TS based upon 
Criteria 1 through 4 of the ``Final Policy Statement on Technical 
Specifications Improvements for Nuclear Power Reactors,'' dated July 
22, 1993. Controls and limitations for testing the refueling system 
interlocks will be incorporated into the USAR. The requirements for 
testing refueling system interlocks are already contained in Station 
procedures. This revision makes the FCS Technical Specifications 
more similar to STS, which do not contain requirements or 
restrictions pertaining to testing refueling system interlocks.
    The proposed revision to TS 3.2, Table 3-5, Item 10, ensures 
consistent use of terminology among the frequencies specified in 
Table 3-5. The proposed revision clarifies the wording and 
introduces additional operational flexibility such that the 
surveillance could be performed before 720 hours of system 
operation, if warranted by plant conditions or beneficial to plant 
operation. Therefore, the proposed revision does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The remaining TS revisions are administrative in nature in that 
they correct references, titles, misspelling(s), and page numbers, 
or revise wording to be consistent with defined intervals within the 
TS. Therefore, they do not increase the probability or consequences 
of an accident previously evaluated. None of the proposed TS 
revisions will impact the function or method of operation of plant 
systems, structures, or components.
    (2) The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed revisions to TS 5.5 and 5.8 which delete the review 
and/or audit of the emergency, site security and safeguards 
contingency plans and implementing procedures from the TS are 
administrative in nature and in accordance with the guidance of GL 
93-07. The proposed revisions will not affect the operation of any 
system, structure, or component and therefore do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed revision to TS 2.5 will still require a backup 
supply of water for the emergency feedwater storage tank to be 
available. However, several other available sources of water are 
preferred over river water, such as, the water plant demineralized 
water system and the outside condensate storage tank. Therefore, the 
proposed revision does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed deletion of TS 2.8(8) pertaining to fuel handling 
cranes, deletion of TS 2.11 pertaining to overhead cranes in the 
Containment and Auxiliary Buildings and deletion of statements in 
the bases of TS 2.8 pertaining to crane interlocks does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated. Specifications 2.8(8), 2.11 and the 
deleted statements in the bases of Specification 2.8 need not be 
retained in the TS based upon Criteria 1 through 4 of the ``Final 
Policy Statement on Technical Specifications Improvements for 
Nuclear Power Reactors,'' dated July 22, 1993.
    The requirements of TS 2.8(8) and restrictions of TS 2.11 are 
currently contained in Station procedures to ensure that the 
handling of fuel assemblies, CEAs and heavy loads is accomplished 
safely and effectively. These revisions make the FCS Technical 
Specifications more similar to STS, which do not contain 
requirements or restrictions concerning the operation of fuel 
handling cranes or overhead cranes.
    The proposed revision to TS 3.2, Table 3-5, Item 1, is an 
administrative revision to the frequency of CEA drop time testing. 
The proposed frequency is the most appropriate time to perform the 
surveillance to ensure that the CEAs drop into the core within the 
time specified in safety analysis and is identical to the frequency 
specified in STS 3.1.5.7. Therefore, the proposed revision does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    The proposed deletion of TS 3.2, Table 3-5, Item 5, which 
currently requires testing the refueling system interlocks prior to 
the refueling outage, does not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
Table 3-5, Item 5, does not need to be retained in the TS based upon 
Criteria 1 through 4 of the ``Final Policy Statement on Technical 
Specifications Improvements for Nuclear Power Reactors,'' dated July 
22, 1993. The requirements for testing refueling 
[[Page 18628]] system interlocks are currently contained in Station 
procedures. This revision makes the FCS Technical Specifications 
more similar to STS, which do not contain requirements or 
restrictions pertaining to testing refueling system interlocks.
    The proposed revision to TS 3.2, Table 3-5, Item 10, ensures 
consistent use of terminology among the frequencies specified in 
Table 3-5. The proposed revision clarifies the wording and 
introduces additional operational flexibility such that the 
surveillance could be performed before 720 hours of system 
operation, if warranted by plant conditions or beneficial to plant 
operation. Therefore, the proposed revision does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    The remaining TS revisions are administrative in nature in that 
they correct references, titles, misspelling(s), and page numbers, 
or revise wording to be consistent with defined intervals within the 
TS. Therefore, they do not create the possibility of a new or 
different kind of accident.
    (3) The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed revisions to TS 5.5 and 5.8 concerning the review 
and/or audit of the emergency, site security and safeguards 
contingency plans and implementing procedures do not involve a 
significant reduction in a margin of safety. The audit and review 
processes are administrative functions which will be retained 
outside the TS in a manner that fully satisfies regulatory 
requirements.
    Removing the requirement of TS 2.5 that Missouri River water 
from the fire water system shall be available to provide a backup 
water supply to the emergency feedwater storage tank improves 
operational flexibility without reducing any safety margins. Better 
sources of backup water are available to replenish the emergency 
feedwater storage tank. Although deleted from TS 2.5, the fire water 
system is still required to be available to meet the requirements of 
paragraph 3.F of the FCS Operating License. Therefore, the proposed 
revision does not involve a significant reduction in a margin of 
safety.
    The proposed deletion of TS 2.8(8) pertaining to fuel handling 
cranes, deletion of TS 2.11 pertaining to overhead cranes in the 
Containment and Auxiliary Buildings and deletion of statements in 
the bases of TS 2.8 pertaining to crane interlocks does not involve 
a significant reduction in a margin of safety. Specifications 
2.8(8), 2.11 and the deleted statements in the bases of 
Specification 2.8 do not need to be retained in the TS based upon 
Criteria 1 through 4 of the ``Final Policy Statement on Technical 
Specifications Improvements for Nuclear Power Reactors,'' dated July 
22, 1993.
    The requirements of Specification 2.8(8) and restrictions of 
Specification 2.11 are currently contained in Station procedures to 
ensure that the handling of fuel assemblies, CEAs and heavy loads is 
accomplished safely and effectively. These revisions make the FCS 
Technical Specifications more similar to STS, which do not contain 
requirements or restrictions concerning the operation of fuel 
handling cranes or overhead cranes.
    The proposed revision to TS 3.2, Table 3-5, Item 1, is an 
administrative revision to the frequency of CEA drop time testing. 
The proposed frequency is the most appropriate time to perform the 
surveillance to ensure that the CEAs drop into the core within the 
time specified in the safety analysis and is identical to the 
frequency specified in STS 3.1.5.7. Therefore, the proposed revision 
does not involve a significant reduction in a margin of safety.
    The proposed deletion of TS 3.2, Table 3-5, Item 5, which 
currently requires testing the refueling system interlocks prior to 
the refueling outage does not involve a significant reduction in a 
margin of safety. Table 3-5, Item 5, does not need to be retained in 
the TS based upon Criteria 1 through 4 of the ``Final Policy 
Statement on Technical Specifications Improvements for Nuclear Power 
Reactors,'' dated July 22, 1993. The requirements for testing 
refueling system interlocks are currently contained in Station 
procedures. This revision makes the FCS Technical Specifications 
more similar to STS, which do not contain requirements or 
restrictions pertaining to testing refueling system interlocks.
    The proposed revision to TS 3.2, Table 3-5, Item 10, ensures 
consistent use of terminology among the frequencies specified in 
Table 3-5. The proposed revision clarifies the wording and 
introduces additional operational flexibility such that the 
surveillance could be performed before 720 hours of system operation 
if warranted by plant conditions or beneficial to plant operation. 
Therefore, the proposed revision does not involve a significant 
reduction in a margin of safety.
    The remaining TS revisions are administrative in nature in that 
they correct references, titles, misspelling(s), and page numbers, 
or revise wording to be consistent with defined intervals within the 
TS. Therefore, they 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: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102
    Attorney for licensee: LeBoeuf, Lamb, Leiby, and MacRae, 1875 
Connecticut Avenue, NW., Washington, DC 20009-5728
    NRC Project Director: William H. Bateman

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 
50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston 
County, Alabama

    Date of amendments request: March 6, 1995
    Description of amendments request: The proposed amendment would 
relocate the seismic and meteorological monitoring instrumentation from 
the Technical Specifications to the Final Safety Analysis Report in 
accordance with the ``Final Policy Statement on Technical 
Specifications Improvements for Nuclear Power Reactors,'' dated July 
22, 1993.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed change relocates information from the TS to the 
FSAR and has no impact on physical plant operation or configuration. 
The continued capability of the seismic and meteorological 
instrumentation to perform its intended function will be ensured 
through controlled change processes governed by 10 CFR 50.59.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. The sole function of the seismic and meteorological 
monitoring instrumentation is to record data. The proposed change 
will not involve any design change or modification to the plant. The 
proposed change will not alter the operation of the plant or the 
manner in which it is operated. Any subsequent change to the Seismic 
and Meteorological Monitoring Instrumentation requirements will 
undergo a review in accordance with the criteria of 10 CFR 50.59 to 
endure that the change does not involve an unreviewed safety 
question.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. The proposed change will relocate Seismic and Meteorological 
Monitoring Instrumentation requirements from the TS to licensee 
controlled documents subject to the criteria of 10 CFR 50.59. The 
proposed change will have no adverse impact on any protective 
boundary or safety limit.
    Therefore, the proposed change will not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Houston-Love Memorial 
[[Page 18629]] Library, 212 W. Burdeshaw Street, Post Office Box 1369, 
Dothan, Alabama 36302
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201
    NRC Project Director: William H. Bateman

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

    Date of amendment request: November 15, 1994; superseded March 7, 
1995 (TS 94-12).
    Description of amendment request: The proposed change would remove 
the frequency for each of the audits specified in the administrative 
controls section of the technical specifications (TS), except those 
related to the fire protection system. The requirements to perform the 
audits would be retained, but the frequency for their performance would 
be controlled by a requirement to be added to the Nuclear Quality 
Assurance Plan. This would require that the audits listed in the TS 
(except those related to the fire protection system) be performed on a 
biennial frequency. In addition, the proposed change would remove the 
requirement to perform site Radiological Emergency Plan, Physical 
Security Plan, and the Safeguard Contingency Plan reviews and audits 
from the TS, since these requirements presently exist in their 
respective Plans.
    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 standards used to arrive at a determination that a Technical 
Specification change request involves no significant hazards 
consideration are included in the Commission's regulations, 10 CFR 
50.92, which states that no significant hazards considerations are 
involved if the 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. Each standard is addressed as 
follows:
    1. Operation of the facility in accordance with the proposed 
technical specifications would not 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 the Technical Specification change which only 
affects review and audit frequencies. This Technical Specification 
change 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 this change. No 
systems, equipment, or components are affected by the proposed 
changes. Thus, the consequences of a malfunction of equipment 
important to safety previously evaluated in the FSAR are not 
increased by this change.
    The proposed change only affects review and audit frequencies. 
As such, the proposed change has no impact on accident initiators or 
plant equipment, and thus, does not affect the probabilities or 
consequences of an accident.
    Therefore, we conclude that this change does not significantly 
increase the probabilities or consequences of an accident.
    2. Operation of the facility in accordance with the proposed 
technical specifications 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 program audits do not contribute to 
accident initiation, a change related to audit functions cannot 
produce a new accident scenario or produce a new type of equipment 
malfunction. Also, this change does not alter any existing accident 
scenarios. The proposed change does not affect equipment or its 
operation, and, thus, does not create the possibility of a new or 
different kind of accident. Therefore, the proposed change does not 
create the possibility of a new or different kind of accident.
    3. Operation of the facility in accordance with the proposed 
technical specifications would not involve a significant reduction 
in a margin of safety.
    The proposed change concerning conduct of reviews and audits 
does not directly affect plant equipment or operation. Safety limits 
and limiting safety system settings are not affected by this 
proposed change.
    Therefore, use of the proposed Technical Specification would not 
involve any reduction in the margin of safety.
    The NRC has reviewed the licensee's analysis and, based on 
thisreview, 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. 50-338 and 50-339, 
North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of amendment request: March 2, 1995
    Description of amendment request: The proposed changes would revise 
Technical Specification 4.6.1.2.a to reference the testing requirements 
of 10 CFR Part 50, Appendix J, and to state that the Nuclear Regulatory 
Commission-approved exemptions to the applicable regulatory 
requirements are permitted.
    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 discussion of these standards as they relate to this ... 
amendment request follows.
    Criterion 1 - Does Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated.
    The proposed change ... revises the North Anna Units 1 and 2 
Technical Specification Surveillance Requirement 4.6.1.2.a to 
reference the testing frequency requirements of 10 CFR 50 Appendix J 
and to state that NRC approved exemptions to the applicable 
regulatory requirements are permitted. The current Technical 
Specification requires Type A tests be conducted in accordance with 
Appendix J to 10 CFR 50. The proposed administrative change simply 
includes the statement ``as modified by NRC-approved exemptions.'' 
No new requirements are added, nor are any existing requirements 
deleted. Any specific changes to the requirements of Appendix J will 
require a submittal from Virginia Electric and Power Company under 
10 CFR 50.12 and subsequent review and approval by the NRC prior to 
implementation. The proposed change is stated generically to avoid 
the need for further Technical Specification changes if different 
exemptions are approved in the future.
    The proposed change, in itself, does not affect reactor 
operations or accident analyses and has no radiological 
consequences. The change provides clarification so that future 
Technical Specifications changes will not be necessary to correspond 
to applicable NRC-approved exemptions from the requirements of 
Appendix J. This exemption request is consistent with the intent of 
the regulation.
    Therefore, this proposed change does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    Criterion 2 - Does Not Create the Possibility of a New or 
Different Kind of Accident from any Previously Evaluated.
    The proposed Technical Specification amendment for Units 1 and 2 
provides clarification to a specification that paraphrases a 
codified requirement.
    Since the ... proposed Technical Specifications change would not 
change the [[Page 18630]] design, configuration, or method of 
operation of the plant, the changes would not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    Criterion 3 - Does Not Involve a Significant Reduction in the 
Margin of Safety.
    The proposed North Anna Units 1 and 2 Technical Specifications 
change is administrative and clarifies the relationship between the 
requirements of Technical Specification Surveillance Requirement 
4.6.1.2.a, Appendix J, and any approved exemptions to Appendix J. It 
does not, in itself, change a Safety Limit or a Limiting Condition 
for Operation. The NRC will directly approve any proposed change or 
exemption to Appendix J prior to implementation.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Project Director: David B. Matthews

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: November 10, 1994
    Description of amendment request: The proposed amendment request 
will clarify the surveillance requirements for the reactor protection 
and the engineered safeguards system instrumentation and actuation 
logic.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Operation of Surry Power Station in accordance with the proposed 
Technical Specifications change will not:
    1. Involve a significant increase in the probability of 
occurrence or consequences of an accident previously evaluated.
    The proposed change to clarify the surveillance requirements for 
the Reactor Protection and Engineered Safeguards Systems 
instrumentation and actuation logic has no impact on the probability 
of an accident occurrence. The instrumentation and actuation logic 
will continue to be operated in the same manner. The actual test 
frequency is not changing. Rather, surveillance requirements are 
being clarified to represent the actual testing and the licensing 
and design bases. Testing of these instruments and actuation logic 
are presently design limited and would otherwise require using 
temporary modifications to complete the testing. Since the testing 
is not changing, the clarification of the actual testing does not 
contribute to the probability of any previously analyzed accident. 
The Reactor Protection and Engineered Safeguards Systems 
instrumentation and actuation logic will be operated in the same 
manner and the system operability requirements are not being 
altered. Therefore, the consequences of any design basis accident 
are not being increased by the proposed change to clarify the 
surveillance test requirements for the Reactor Protection and 
Engineered Safeguards System instrumentation and actuation logic.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    There are no plant modifications or changes in methods of plant 
operation introduced by this change in the clarification of the 
testing for the Reactor Protection and Engineered Safeguards Systems 
instrumentation and actuation logic. The plant is not being operated 
or tested in a different manner due to the proposed change. 
Therefore, no new accidents or accident precursors are generated by 
the proposed change to clarify the surveillance test requirements.
    Clarifying the surveillance test requirements to represent the 
original licensing design basis and test conditions does not create 
the possibility of a new or different accident than previously 
analyzed.
    3.Involve a significant reduction in a margin of safety.
    Clarification of the testing for the Reactor Protection and 
Engineered Safeguards Systems instrumentation and actuation logic 
surveillance requirements does not affect the margin of safety in 
that the operability requirements for these safety systems remain 
unchanged. The existing testing is performed in accordance with 
plant design and licensing basis and provides adequate indication of 
the operability of the affected instrumentation or actuation logic. 
The Reactor Protection and Engineered Safeguards Systems 
instrumentation and actuation logic are fully tested on a refueling 
cycle basis which includes complete operation of each relay and end 
device. Therefore, the margin of safety is not altered by the 
proposed clarification of the testing for the Reactor Protection and 
Engineered Safeguards Systems instrumentation and actuation logic.
    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: David B. Matthews

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: November 22, 1994
    Description of amendment request: The proposed amendment request 
would delete unnecessary descriptive phrases regarding the number of 
cells in the station and emergency diesel generator batteries.
    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 deletion of the descriptive references regarding the number 
of cells in the station and emergency diesel generator batteries is 
an administrative change and therefore does not:
    1. Involve an increase in the probability of occurrence or 
consequences of an accident previously evaluated.
    The proposed change to delete the descriptive references 
associated with the station and emergency diesel generator batteries 
(60 cell or 56 cell, respectively) has no impact on the probability 
of an accident occurrence. The change is administrative in nature 
and therefore does not affect the operation of the units. The 
batteries will continue to be operated in the same manner as before 
the change with operability based on design voltage and capacity 
requirements necessary to ensure safety functions can be performed. 
Prescribed surveillance testing will continue to ensure the 
operability of individual battery cells. Consequently, the proposed 
change does not contribute to the probability of occurrence or 
consequences of any design basis accident.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    This is an administrative change to delete the descriptive 
references associated with the station and emergency diesel 
generator batteries. There are no plant modifications being 
implemented by the proposed change and plant operations are not 
being changed. Provided the required design voltage and capacity are 
maintained, the batteries remain fully operable and capable of 
performing their intended safety functions. Individual battery cell 
surveillance requirements remain unchanged. Therefore, no new 
accidents or accident precursors are created by the proposed change.
    3. Involve a reduction in a margin of safety as defined in the 
Technical Specifications. [[Page 18631]] 
    The proposed administrative change to delete the descriptive 
references associated with the station and emergency diesel 
generator batteries (60 cell or 56 cell, respectively) is 
administrative in nature. Provided the required design voltage and 
capacity are maintained, the batteries remain fully operable and 
capable of performing their intended safety functions as assumed in 
the safety analyses. Individual battery cell surveillance 
requirements remain unchanged. Therefore, the analyzed margin of 
safety is not reduced by the proposed change.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: 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: David B. Matthews

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: January 24, 1995
    Description of amendment request: The proposed amendment request 
would increase the current Technical Specification pressurizer safety 
valve lift setpoint acceptance criterion from plus or minus 1% as-found 
and plus or minus 1% as-left to plus or minus 3% as-found and plus or 
minus 1% as-left.
    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 Technical Specifications change does not involve a 
significant hazards consideration because operation of Surry Units 1 
and 2 in accordance with this change would not:
    a. involve a significant increase in the probability or 
consequences of an accident previously evaluated. Affected safety-
related parameters were analyzed for a change to Surry Units 1 and 2 
Technical Specification 3.1.A.3.b. It was determined that the 
primary and secondary side overpressure safety limits would not be 
exceeded in the most limiting overpressure transient (Loss of Load, 
Locker Rotor, and Rod Withdrawal events) with the pressurizer safety 
valve lift setpoint acceptance criterion increased to [plus or 
minus] 3%. The DNBR [departure from nucleate boiling ratio] results 
of transients impacted by the setpoint acceptance criterion increase 
are not affected by the proposed change. The increased setpoint 
acceptance criterion will not result in an inadvertent opening of 
the pressurizer safety valves. Since the proposed change involves no 
alterations to the physical plant, the probability of occurrence of 
an accident or malfunction of equipment important to safety 
previously evaluated is not increased.
    b. create the possibility of a new or different kind of accident 
from any accident previously identified. The proposed change to 
Surry Units 1 and 2 Technical Specification 3.1.A.3.b does not 
involve any alterations to the physical plant which would introduce 
any new or unique operational modes or accident precursors. Only the 
allowable tolerance about the existing setpoint will be changed.
    c. involve a significant reduction in a margin of safety. It was 
determined that the most limiting overpressure transients do not 
result in maximum pressures in excess of the primary and secondary 
side overpressure limits. The DNBR results of affected transients 
are not made more limiting by the proposed setpoint tolerance 
increase. Therefore, the margin of safety is unchanged by the 
proposed increase in the safety valve setpoint acceptance criterion.
    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: David B. Matthews

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf 
Creek Generating Station, Coffey County, Kansas

    Date of amendment request: March 21, 1995
    Description of amendment request: The amendment would revise 
Surveillance Requirement 4.6.2.1.d for the containment spray system to 
change the surveillance interval for the performance of the air or 
smoke test through the containment spray header from once per 5 years 
to once per 10 years.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed reduced testing frequency of the Containment Spray 
System nozzles does not change the way the system is operated or the 
Containment Spray System's operability requirements. The proposed 
change to the surveillance frequency of safety equipment has no 
impact on the probability of an accident occurrence nor can it 
create a new or different type of accident. NUREG-1366 concluded 
that the corrosion of stainless steel piping is negligible during 
the extended surveillance interval. Since the Containment Spray 
System is maintained dry there is no additional mechanism that could 
cause blockage of the spray nozzles. Thus, the nozzles in the 
Containment Spray System will remain operable during the ten year 
surveillance interval to mitigate the consequence of an accident 
previously evaluated. No clogging or blockage of the nozzles in the 
Containment Spray System has been discovered during the performance 
of the five year surveillance tests. Therefore, the testing of the 
Containment Spray System[']s nozzles at the proposed reduced 
frequency will not increase the probability or consequences of an 
accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed reduced frequency testing of the Containment Spray 
System nozzles does not change the way the Containment Spray System 
is operated. The reduced frequency of testing of the spray nozzles 
does not change plant operation or system readiness. The reduced 
frequency testing of the Containment Spray System nozzles does not 
generate any new accident precursors. Therefore, the possibility of 
a new or different kind of accident from any accident previously 
evaluated is not created by the proposed changes in surveillance 
frequency of the Containment Spray System nozzles.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    Reduced testing of the Containment Spray System nozzles does not 
change the way the system is operated or the Containment Spray 
System's operability requirements. NUREG-1366 concluded that the 
corrosion of stainless steel piping is negligible during the 
extended surveillance interval. Since the Containment Spray System 
is maintained dry there is no additional mechanism that could cause 
blockage of the Containment Spray System nozzles. Thus, the proposed 
reduced testing frequency is adequate to ensure spray nozzle 
operability. The surveillance requirements do not affect the margin 
of safety in the operability requirements of the Containment Spray 
System remains unaltered. The existing safety analysis remains 
bounding. Therefore no margins of safety are adversely affected by 
this proposed change.
    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 
[[Page 18632]] satisfied. Therefore, the NRC staff proposes to 
determine that the amendment request involves no significant hazards 
consideration.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
    NRC Project Director: William H. Bateman

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf 
Creek Generating Station, Coffey County, Kansas

    Date of amendment request: March 24, 1995
    Description of amendment request: The proposed amendment would add 
a new action statement to Technical Specification 3.5.1 which would 
provide a 72-hour allowed outage time (AOT) for one accumulator to be 
inoperable because its boron concentration did not meet the 2300-2500 
parts per million (ppm) band. The amendment would also change the 
current allowed outage time for other reasons of inoperability from 1 
hour to 24 hours.
    Changes to the surveillance requirements are also proposed to 
incorporate the guidance of Generic Letter 93-05, ``Line-Item Technical 
Specifications Improvements to Reduce Surveillance Requirements for 
Testing During Operation.'' These proposed changes would base the 
operability of the accumulator on the contained water volume and cover 
pressure and would not require verification of the boron concentration 
after an accumulator volume increase, provided the source of the makeup 
water is the refueling water storage tank.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) The proposed change does not involve a significant Increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
The overall protection system performance will remain within the 
bounds of the accident analysis documented in Chapter 15 of the 
Updated Safety Analysis Report [USAR], WCAP-1096-P, and WCAP-11883 
since no hardware changes are proposed.
    The safety injection accumulators are credited in Section 15.6.5 
of the Updated Safety Analysis Report for large and small break LOCA 
[loss-of-coolant accident]. There will be no effect on these 
analyses, or any other accident analysis, since the analysis 
assumptions are unaffected and remain the same as discussed in 
Section 15.6.5. Design basis accidents are not assumed to occur 
during allowed outage times covered by the Technical Specifications. 
As such, the ECCS [emergency core cooling system] Evaluation Model 
equipment availability assumptions made in Section 15.6.5 remain 
valid.
    The safety injection accumulators will continue to function in a 
manner consistent with the above analysis assumptions and the plant 
design basis. As such, there will be no degradation in the 
performance of nor an increase in the number of challenges to 
equipment assumed to function during an accident situation.
    The proposed technical specifications changes do not involve any 
hardware changes nor do they affect the probability of any event 
initiators. There will be no change to normal plant operating 
parameters, ESF [engineered safety features] actuation setpoints, 
accident mitigation capabilities, accident analysis assumptions or 
inputs. Therefore, these changes will not increase the probability 
of an accident or malfunction.
    The corresponding increase in CDF [core damage frequency] due to 
the proposed change to increase the AOT of the accumulators from one 
hour to 24 hours is insignificant. Pursuant to the guidance in 
Section 3.5 of NSAC-125, the proposed increase in AOT does not 
``degrade below the design basis the performance of a safety system 
assumed to function in the accident analysis,'' nor does it 
``increase challenges to safety systems assumed to function in the 
accident analysis such that safety system performance is degraded 
below the design basis without compensating effects.'' Therefore, it 
is concluded that these changes do not increase the probability of 
occurrence of a malfunction of equipment important to safety.
    (2) The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
This change is administrative in nature and does not involve any 
change to the installed plant systems or the overall operating 
philosophy of WCGS [Wolf Creek Generating Station].
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of these proposed changes. There will be no adverse effect or 
challenges imposed on any safety-related system as a result of these 
changes. Therefore, the possibility of a new or different type of 
accident is not created.
    There are no changes which would cause the malfunction of 
safety-related equipment, assumed to be operable in the accident 
analyses, as a result of the proposed technical specification 
changes. No new mode failure has been created and no new equipment 
performance burdens are imposed. Therefore, the possibility of a new 
or different malfunction of safety-related equipment is not created.
    (3) The proposed change does not involve a significant reduction 
in the margin of safety.
    The proposed change does not involve an significant reduction in 
a margin of safety. There will be no change to the Departure from 
Nucleate Boiling Ratio (DNBR) Correlation Limit, the design DNBR 
limits, or the safety analysis DNBR limits discussed in Bases 
Section 2.1.1.
    As discussed previously, the performance of the accumulators 
will remain within the assumptions used in the large and small break 
LOCA analyses, as presented in USAR Section 15.6.5. Also, there will 
be no effect on the manner in which safety limits or limiting safety 
system settings are determined nor will there be any effect on those 
plant systems necessary to assure the accomplishment of protection 
functions.
    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 locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
    NRC Project Director: William H. Bateman

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. [[Page 18633]] 

Baltimore Gas and Electric Company, Docket No. 50-318, Calvert 
Cliffs Nuclear Power Plant, Unit No. 2, Calvert County, Maryland

    Date of amendment request: February 24, 1995
    Brief description of amendments: The proposed amendment would 
revise the Calvert Cliffs, Unit No. 2, Technical Specifications (TSs). 
Specifically, TS 4.G.1.2 would reference 10 CFR Part 50, Appendix J, 
directly, and any approved exemptions to the Type A testing frequency 
requirements, rather than paraphrase the regulation. The proposed 
wording is consistent with that used in NUREG-1432, ``Standard 
Technical Specifications - Combustion Engineering Plants,'' dated 
September 1992.Date of publication of individual notice in Federal 
Register: March 8, 1995 (60 FR 12789)
    Expiration date of individual notice: April 7, 1995
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

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

    Date of amendment request: February 23, 1995, as supplemented March 
21, 1995.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications 3.8.2.1 and 3.8.3.1 to allow 
installation of a modification to replace the battery, main and tie 
breakers in response to an Electrical Distribution Systems Functional 
Inspection, conducted by the NRC in July 1991. The existing breaker 
arrangement could result in a trip of both the battery and main 
breakers if a fault occurs on one of the 125 VDC panelboards. The 
licensee committed to have these breakers replaced in 1995 with a 
better coordinated design to eliminate the concern.Date of publication 
of individual notice in Federal Register: March 8, 1995 (60 FR 12791)
    Expiration date of individual notice: April 7, 1995
    Local Public Document Room location: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina.

Houston Lighting & Power Company, City Public Service Board of San 
Antonio, Central Power and Light Company, City of Austin, Texas, 
Docket No. 50-498, South Texas Project, Unit 1, Matagorda County, 
Texas

    Date of amendment request: March 1, 1995
    Description of amendment request: The proposed amendment would 
modify the steam generator tube plugging criteria in Technical 
Specification 3/4.4.5, Steam Generators, and the allowable leakage for 
Unit 1 in Technical Specification 3/4.4.6.2, Operational Leakage, and 
the associated Bases.Date of individual notice in the Federal Register: 
March 13, 1995 (60 FR 13478)
    Expiration date of individual notice: April 12, 1995
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488

Houston Lighting & Power Company, City Public Service Board of San 
Antonio, Central Power and Light Company, City of Austin, Texas, 
Docket No. 50-498, South Texas Project, Unit 1, Matagorda County, 
Texas

    Date of amendment request: March 1, 1995
    Description of amendment request: The proposed amendment would 
change Technical Specification 3/4.4.5, Steam Generators, and the 
associated Bases to allow the use of an alternate plugging criteria 
(known in the industry as F*) on steam generator tubes that are 
defective or degraded within certain areas within the tubesheet. Date 
of individual notice in the Federal Register: March 13, 1995 (60 FR 
13481)
    Expiration date of individual notice: April 12, 1995
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile 
Point Nuclear Station, Unit 2, Oswego County, New York

    Date of amendment request: March 9, 1994
    Description of amendment request: The proposed amendment would 
revise the Nine Mile Point Nuclear Station, Unit 2, Technical 
Specifications (TSs). Specifically, TS 4.6.1.2.a would be modified to 
allow the second Primary Containment Integrated Leakage Rate Test (Type 
A) to be performed at the fifth refueling outage (RF-05) or 72 months 
after the first Type A test instead of the fourth refueling outage (RF-
04) as currently scheduled.
    Date of publication of individual notice in Federal Register: March 
23, 1995 (60 FR 15310)
    Expiration date of individual notice:  April 24, 1995
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. 
Ginna Nuclear Power Plant, Rochester, New York

    Date of application for amendment: March 13, 1995
    Brief description of amendment: The proposed amendment would revise 
Ginna Station Technical Specification (TS) 4.4.2.4.a to replace 
specific leakage testing frequencies for containment isolation valves. 
This TS change will support a proposed Exemption to Title 10 of the 
Code of Federal Regulations (10 CFR) Part 50, Appendix J, Section 
III.D.3, requested under separate cover to exempt Type C testing of 
certain valves during a 1995 refueling outage.
    Date of publication of individual notice in Federal Register: March 
22, 1995 (60 FR 15167)
    Expiration date of individual notice: April 21, 1995
    Local Public Document Room location: Rochester Public Library, 115 
South Avenue, Rochester, New York 14610.

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 [[Page 18634]] under the special circumstances 
provision in 10 CFR 51.12(b) and has made a determination based on that 
assessment, it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document rooms for 
the particular facilities involved.

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power 
Station,Plymouth County, MassachusettsDate of application for 
amendment: November 22, 1994

    Brief description of amendment: The amendment revises the allowable 
leak rate for the main steam isolation valves from the current 11.5 
standard cubic feet per hour (scfh) for each valve, to a maximum 
combined main steam line leak rate of 46 scfh.
    Date of issuance: March 22, 1995
    Effective date: March 22, 1995
    Amendment No.: 160
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 18, 1995 (60 FR 
3671) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 22, 1995. No significant hazards 
consideration comments received: No
    Local Public Document Room location: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power 
Station,Plymouth County, Massachusetts

    Date of application for amendment: September 6, 1994, as 
supplemented February 15, 1995.
    Brief description of amendment: This amendment revises Technical 
Specifications (TSs) 3.7.B.1.a, 3.7.B.1.c, 3.7.B.1.e, 3.7.B.2.a, and 
3.7.B.2.c and adds Sections 3.7.B.1.f and 3.7.B.2.e. The additional 
section requires both trains of standby gas treatment and control room 
high efficiency air filtration system to be operable for the initiation 
of fuel movement. In the event either train becomes inoperable, the 
other train must be demonstrated to be operable within 2 hours and fuel 
handling operations may continue for 7 days with one train inoperable. 
Additionally, this change allows one train to be defined as operable 
without its associated emergency power supply, provided one source of 
normal power (startup transformer or unit auxiliary power) is 
available.
    Date of issuance: March 22, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 161
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 26, 1994 (59 FR 
53837) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 22, 1995. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power 
Station,Plymouth County, Massachusetts

    Date of application for amendment: September 6, 1994
    Brief description of amendment: This amendment would reduce the 
Reactor Pressure Setpoint at which the shutdown cooling system 
automatically isolates. This setpoint also isolates the low pressure 
coolant injection valves when the shutdown cooling system is in 
operation.
    Date of issuance: March 27, 1995
    Effective date: To be implemented within 30 days following restart 
from refueling outage <greek-i>10
    Amendment No.: 162
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 26, 1994 (59 FR 
53837) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 27, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.

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 application for amendments: October 28, 1994, as 
supplemented February 16, 1995.
    Brief description of amendments: The proposed change will revise TS 
requirements to increase the surveillance test intervals and the 
allowable out of service times or instruments of the reactor protection 
system, isolation actuation system, emergency core cooling system 
actuation system, control rod withdrawal block system, control room 
emergency ventilation system, anticipated transient without scram, 
recirculation pump trip (RPT), end-of-cycle RPT, and the reactor core 
isolation cooling actuation system.
    Date of issuance: March 30, 1995Effective date: March 30, 1995
    Amendment Nos.: 175 and 206
    Facility Operating License Nos. DPR-71 and DPR-62. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 7, 1994 (59 FR 
63114) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 30, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, 
North CarolinaDate of application for amendment: October 24, 1994, 
as supplemented December 6, 1994.

    Brief description of amendment: The amendment allows the relocation 
of TS 3/4.3.4, Turbine Overspeed Protection and associated Bases to be 
consistent with the new Standard Technical Specifications for 
Westinghouse plants.
    Date of issuance: March 22, 1995
    Effective date: March 22, 1995
    Amendment No. 55
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications
    Date of initial notice in Federal Register: November 23, 1994 (59 
FR 60379) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 22, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 
27605. [[Page 18635]] 

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, IllinoisDocket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, 
Will County, Illinois

    Date of application for amendments: December 22, 1992
    Brief description of amendments: These amendments add new 
requirements to the Technical Specifications (TS) to ensure that an 
Essential Service Water system (SX) pump and crossover path are 
available from a shutdown unit to serve as backup to an operating unit. 
In addition, a new TS is added to require the unit crosstie to be open, 
or capable of being opened, from the Main Control Room, whenever 
either, or both units are in an operating mode (MODE 1, 2, 3, or 4).
    Date of issuance: March 20, 1995
    Effective date: March 20, 1995
    Amendment Nos.: 71, 71, 62, and 62
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 3, 1993 (58 FR 
6994) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 20, 1995. No significant 
hazards consideration comments received: No
    Local Public Document Room location: For Byron, the Byron Public 
Library, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for 
Braidwood, the Wilmington Township Public Library, 201 S. Kankakee 
Street, Wilmington, Illinois 60481.

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: September 19, 1994
    Brief description of amendment: The amendment would revise 
Technical Specification Section 4.4.A.3, Frequency of Containment 
Integrated Leakage Rate Test, to reference 10 CFR Part 50, Appendix J, 
as modified by approved exemptions, directly.
    Date of issuance: March 17, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment No.: 181
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 15, 1995 (60 
FR 8744) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 17, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, 
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport, 
PennsylvaniaDate of application for amendments: April 23, 1990, as 
supplemented January 21, 1992 and March 17, 1995.

    Brief description of amendments: These amendments revise the 
Appendix A Technical Specifications (TSs) for Unit 1 and Unit 2 by (a) 
deleting TS Table 3.6-1, ``Containment Penetrations,'' (b) rewording TS 
Definition 1.8, ``Containment Integrity,'' and TSs 3.6.1.1, 3.6.1.2, 
3.6.3.1, and 3.9.4 relating to containment integrity, containment 
leakage, containment isolation valves, and containment building 
penetrations respectively to account for the deletion of TS Table 3.6-
1, and (c) correcting terminology by replacing the word ``door'' with 
``hatch'' in TS 3.9.4.a.
    The Unit 1 amendment also modifies TS Table 3.3-5, ``Engineered 
Safety Features Response Times,'' by changing the feedwater isolation 
response time to reflect total isolation times for the main feedwater 
regulating valve and bypass feedwater regulating valve. Minor editorial 
changes were also incorporated in TS Table 3.3-5.
    Date of issuance: March 28, 1995
    Effective date: March 28, 1995
    Amendment Nos.: 185 and 66
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 27, 1990 (55 FR 
26283), as supplemented April 1, 1992 (57 FR 11107) The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated March 28, 1995. No significant hazards consideration 
comments received: No.
    Local Public Document Room location:  B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.
    GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station, Ocean County, New Jersey
    Date of application for amendment: June 22, 1994
    Brief description of amendment: The amendment changes Technical 
Specification (TS) Sections 1.6, 3.2.A, 3.9.f.5 and 4.2.A which specify 
the Shutdown Margin (SDM) requirements that ensure the reactor can be 
made subcritical and can be maintained sufficiently subcritical to 
preclude inadvertent criticality in any core condition. The amendment 
also includes a definition of Shutdown Margin, TS Section 1.45. 
Administrative changes to TS Sections 1.7 and 3.2.b.2(b) are also 
included to simplify definitions and eliminate unnecessary notes and 
references.
    Date of Issuance: March 21, 1995Effective date: As of the date of 
issuance to be implemented within60 days
    Amendment No.: 178
    Facility Operating License No. DPR-16. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 20, 1994 (59 FR 
37072) The Commission's related evaluation of this amendment is 
contained in a Safety Evaluation dated March 21, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.

Illinois Power Company and Soyland Power Cooperative, Inc., Docket 
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County, 
Illinois

    Date of application for amendment: February 14, 1995
    Brief description of amendment: The amendment revises Technical 
Specification 3.8.2, ``AC Sources-Shutdown;'' 3.8.5, ``DC Sources-
Shutdown;'' and 3.8.8, ``Inverters-Shutdown.'' The changes revise the 
operability requirements for the Division 3 diesel generator and the 
Division 3 and 4 batteries, battery chargers and inverters to apply 
only when the high pressure core spray system is required to be 
operable.
    Date of issuance: March 21, 1995
    Effective date: March 21, 1995
    Amendment No.: 99
    Facility Operating License No. NPF-62. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 17, 1995 (60 
FR 9412) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 21, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: The Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 
61727. [[Page 18636]] 

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

    Date of application for amendment: June 30, 1994, as supplemented 
March 7, 1995
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3.2.7.1 to add 8 check valves to Table 3.2.7.1. 
These valves were installed to add additional protection of the low 
pressure Core Spray system from the high pressure Reactor Coolant 
system. Including the valves in the TSs will assure that the proper 
surveillance testing is done to maintain a high reliability for the 
valves to protect the Core Spray system.
    Date of issuance: March 20, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 154
    Facility Operating License No. DPR-63: Amendment revises 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 March 20, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

Northeast Nuclear Energy Company, et al., Docket No. 50-336, 
Millstone Nuclear Power Station, Unit No. 2, New London County, 
Connecticut

    Date of application for amendment: April 22, 1994
    Brief description of amendment: The amendment deletes the 
operability and surveillance requirements of the condenser air ejector 
radiation monitor from the Millstone Unit 2 Technical Specification 
Tables 3.3-12 and 4.3-12.
    Date of issuance: March 27, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment No.: 186
    Facility Operating License No. DPR-65. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 25, 1994 (59 FR 
27058) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 27, 1995. No significant 
hazards consideration comments received: No.
    Local Public Document Room location:  Learning Resources Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, CT 06360.

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

    Date of application for amendment: March 31, 1994 and August 5, 
1994
    Brief description of amendment: This amendment revises: Technical 
Specification (TS) 3.8.1.1.b.2 which maintains diesel operability for a 
48-hour period when the fuel storage system of one or more diesel 
generators contains less than a 7-day supply of fuel: TS 4.8.1.1.2.h.8 
by deletion and replacement with surveillance requirement 4.8.1.1.2.k.1 
which permits the 24-hour diesel generator endurance run to be 
performed in any operational condition; establish surveillance 
requirement 4.8.1.1.2.k.2 which allows the hot restart test to be 
conducted not only after surveillance requirement 4.8.1.1.2.k.1, but 
also after the diesel generator has operated between 4300 kw and 4400 
kw for one hour or after any time the diesel generator operating 
temperature has stabilized; revise TS 3.8.1.1 to eliminate the 
requirements to start the Emergency Diesel Generator (EDG) with an 
inoperable offsite circuit(s) of AC electrical power; add a provision 
that eliminates required testing of 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, the 
surveillance will be performed within 16 hours instead of 24 hours as 
currently specified; delete the requirement to perform a Loss of 
Offsite Power (LOOP) test (Surveillance Requirement 4.8.1.1.2.h.b) 
following the 24-hour EDG endurance run test in its place, a hot 
restart test (no LOOP load sequencing) will be established.
    Date of issuance: March 30, 1995
    Effective date: March 30,1995
    Amendment No.: 72
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 8, 1994 (59 FR 
29630) and October 12, 1994 (59 FR 51625) The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
March 30, 1995. No significant hazards consideration comments received: 
No
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070

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 application for amendment: October 29, 1993, as 
supplemented on March 11, 1994, May 18, 1994, September 20, 1994, and 
October 20, 1994.
    Brief description of amendment: The amendment changes Operating 
License NPF-12 to delete License Conditions 2.C.13, 2.C.14, and 2.C.32.
    Date of issuance: March 29, 1995
    Effective date: March 29, 1995
    Amendment No.: 123
    Facility Operating License No. NPF-12. Amendment revises the 
operating license.
    Date of initial notice in Federal Register: February 16, 1994 (59 
FR 7698) and April 28, 1994 (59 FR 22012), as corrected June 30, 1994 
(59 FR 33795). The May 18, 1994, September 20, 1994, and October 20, 
1994, submittals provided supplemental and clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated March 29, 1995.No 
significant hazards consideration comments received: No
    Local Public Document Room location: Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180

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: September 30, 1993, as 
supplemented by letters dated November 16, 1993, January 18, 1995, and 
February 2, 1995.
    Brief description of amendments: These amendments revised the 
technical specifications to (1) divide item 7 of Tables 3.3-3, 3.3-4, 
3.3-5, and 4.3-2 into item 7a that addresses the existing loss-of-
voltage (LOV) function and item 7b that separately addresses the 
degraded grid voltage (DGV) function; (2) add footnote (d) to Table 
3.3-3 to indicate that the DGV actuation relay logic is applicable in 
Modes 1, 2, 3, and 4 when the diesel generator circuit breaker is open; 
(3) replace the reference to Figure 3.3-1 in item 7a of Tables 3.3-4 
and 3.3-5 with definite voltage and time values; (4) add note 9 to 
Table 3.3-5 to explain the response [[Page 18637]] time for an LOV 
signal; and (5) delete Figure 3.3-1, ``Degraded Bus Voltage Trip 
Setting,'' and the reference to this figure from Table 3.3-4.
    Date of issuance: March 17, 1995
    Effective date: Unit 2, as of the date of completion of the 
currrent refueling outage and must be fully implemented before the 
plant returns to power; Unit 3, as of the date of the completion of its 
next refueling outage and must be fully implemented before the plant 
returns to power.
    Amendment Nos.: Unit 2 - Amendment No. 118; Unit 3 - Amendment No. 
107
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 10, 1993 (58 
FR 59755). The additional information contained in the November 16, 
1993, January 18, 1995 and February 2, 1995, letters was clarifying in 
nature, within the scope of the initial notice and did not affect the 
NRC staff's proposed no significant hazards consideration 
determination.The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 17, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 
50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston 
County, Alabama.

    Date of amendments request: January 9, 1995
    Brief description of amendments: The amendments change the 
Technical Specifications to implement recommended changes from Generic 
Letter (GL) 93-05, ``Line Item Technical Specification Improvements to 
Reduce Surveillance Requirements for Testing During Power Operation,'' 
dated September 27, 1993. Specifically, the amendments implement TS 
changes corresponding to the following GL 93-05 line-item improvement 
issues and numbers: Control Rod Movement Test for Pressurized Water 
Reactors (4.2.1); Radiation Monitors (5.14); Surveillance of Boron 
Concentration in the Accumulator/Safety Injection/Core Flood Tank 
(7.1); Containment Spray System (8.1); Hydrogen Recombiner (8.5); and 
Special Test Exemptions (12).
    Date of issuance: March 20, 1995
    Effective date: March 20, 1995
    Amendment Nos.: 113 and 104
    Facility Operating License Nos. NPF-2 and NPF-8. Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: February 15, 1995 (60 
FR 8756) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 20, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302

Toledo Edison Company, Centerior Service Company, and The Cleveland 
Electric Illuminating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

    Date of application for amendment: December 6, 1994
    Brief description of amendment: This amendment deletes Technical 
Specification (TS) Surveillance Requirement (SR) 4.1.3.2.2 for the 
Axial Power Shaping Rods and relaxes surveillance intervals for TS 3/
4.1.3.1, ``Group Height - Safety and Regulating Rod Groups;'' TS 3/
4.4.6.2, ``Operational Leakage;'' TS 3/4.5.2, ``ECCS Subsystems - Tavg 
equal to or greater than 280 deg.F;'' TS 3/4.6.2.1, ``Containment Spray 
System;'' and TS 3/4.10.4, ``Special Test Exceptions Shutdown Margin.'' 
Date of issuance: March 21, 1995Effective date: March 21, 1995 and 
implemented not later than 90 days after issuance
    Amendment No.:  196
    Facility Operating License No. NPF-3. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 15, 1995 (60 
FR 8757) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 21, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: University of Toledo Library, 
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.

Toledo Edison Company, Centerior Service Company, and The Cleveland 
Electric Illuminating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

    Date of application for amendment: December 6, 1994
    Brief description of amendment: This amendment revises Technical 
Specification (TS) 4.0.5, ``Applicability'' and its associated Bases to 
eliminate the need for NRC approval of relief requests prior to 
implementation and relaxes surveillance test intervals for TS 3/
4.1.2.3, ``Reactivity Control Systems - Makeup Pump - Shutdown; TS 3/
4.1.2.4, ``Reactivity Control Systems - Makeup Pumps - Operating; TS 3/
4.1.2.6, Reactivity Control Systems - Boric Acid Pump - Shutdown; and 
TS 3/4.1.2.7, ``Reactivity Control System - Boric Acid Pumps - 
Operating'' from monthly to quarterly. Date of issuance: March 22, 1995
    Effective date: March 22, 1995, and to be implemented within 90 
days
    Amendment No.: 197
    Facility Operating License No. NPF-3. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 15, 1995 (60 
FR 8758) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 22, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: University of Toledo Library, 
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application for amendment: December 9, 1994, as 
supplemented on December 22, 1994.
    Brief description of amendment: The amendment revises the Technical 
Specification (TS) Surveillance Requirement 4.8.1.1.2f.7. The change 
removes the requirement to perform the hot restart test within 5 
minutes of completing the 24-hour endurance test and places that 
requirement in a separate TS.
    Date of issuance: March 20, 1995
    Effective date: March 20, 1995, to be implemented within 30 days
    Amendment No.: 95
    Facility Operating License No. NPF-30. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 1, 1995 (60 FR 
6315) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 20, 1995. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251. [[Page 18638]] 

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application for amendment: August 4, 1994, as supplemented 
on March 14, 1995 and March 28, 1995.
    Brief description of amendment: The amendment replaces Technical 
Specification (TS) 3/4.6.2.2, Spray Additive System, with a new TS 3/
4.6.2.2 entitled Recirculation Fluid pH control (RFPC) System. The 
associated TS Surveillance Requirements and the Bases will also be 
revised. In addition, the Bases section for the Refueling Water Storage 
Tank (RWST) System will be revised.
    Date of issuance: March 30, 1995
    Effective date: March 30, 1995, to be implemented within 30 days
    Amendment No.: 96
    Facility Operating License No. NPF-30. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 28, 1994 (59 
FR 49440) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 30, 1995. The March 14, 
1995, and March 28, 1995, letters provided 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: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application for amendment: September 8, 1994
    Brief description of amendment: The amendment revises the Technical 
Specification (TS) Bases Section 3/4.9 and changes Final Safety 
Analysis Report (FSAR) Sections 9.1.3 ``Fuel Pool Cooling and 
Cleanup,'' 9.1.4 ``Fuel Handling System'' and 15.4.6 ``Chemical and 
Volume Control System Malfunction That Results in a Decrease in the 
Boron Concentration in the Reactor Coolant. The changes established 
procedural controls to address an unreviewed safety question.
    Date of issuance: March 31, 1995
    Effective date: March 31, 1995, to be implemented within 30 days
    Amendment No.: 97
    Facility Operating License No. NPF-30. Amendment revises the 
Technical Specification Bases and FSAR.
    Date of initial notice in Federal Register: March 1, 1995 (60 FR 
11151) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 31, 1995. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, 
Vermont Yankee Nuclear Power Station, Vernon, Vermont

    Date of application for amendment: December 8, 1994, as 
supplemented by letter dated February 16, 1995.
    Brief description of amendment: The proposed amendment would change 
Standby Gas Treatment Power Supply Requirements during refueling 
operations.
    Date of issuance: March 23, 1995
    Effective date: As of the date of issuance, to be implemented 
within 30 days
    Amendment No.: 143
    Facility Operating License No. DPR-28. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 15, 1995 (60 
FR 8759) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 23, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, Vermont 05301.

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

    Date of application for amendment: October 31, 1994
    Brief description of amendment: The amendment relocated 
requirements regarding safety/relief valve position indication 
instrumentation from the Technical Specifications to other licensee-
controlled documents.
    Date of issuance: March 27, 1995
    Effective date: March 27, 1995, to be implemented prior to restart 
from the spring 1995 refueling outage
    Amendment No.: 135
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 21, 1994 (59 
FR 65831) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 27, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352

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

    Date of application for amendment: December 2, 1994
    Brief description of amendment: The amendment revises Kewaunee 
Nuclear Power Plant (KNPP) Technical Specification (TS) 3.2 by deleting 
the requirements for the charging pumps, high concentration boric acid 
in the boric acid storage tanks (BASTs), the boric acid transfer pumps, 
and boric acid heat tracing. Changes to TS 3.3 and Table TS 3.5.3 add 
requirements associated with the emergency core cooling system (ECCS) 
accumulators, remove the requirements associated with the boric acid 
storage tanks and increase the minimum required boron concentration in 
the refueling water storage tank (RWST). Additionally, the surveillance 
requirements involving the BASTs, associated valves and heat tracing 
located in Table TS 4.1-1, Table TS 4.1-2 and Section 4.5 have been 
deleted.
    Date of issuance: March 28, 1995
    Effective date: March 28, 1995, to be implemented within 20 days
    Amendment No.: 116
    Facility Operating License No. DPR-43. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 4, 1995 (60 FR 
508). The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 28, 1995. No significant hazards 
consideration comments received: None.
    Local Public Document Room location: University of Wisconsin 
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin 
54301.

Notice Of Issuance Of Amendments To Facility Operating LicensesAnd 
Final Determination Of No Significant Hazards ConsiderationAnd 
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 
[[Page 18639]] 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, 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 May 12, 1995, the licensee 
may file a request for a hearing with respect to issuance of the 
amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should 
consult a current copy of 10 CFR 2.714 which is available at the 
Commission's Public Document Room, the Gelman Building, 2120 L Street, 
NW., Washington, DC and at the local public document room for the 
particular facility involved. If a request for a hearing or petition 
for leave to intervene is filed by the above date, the Commission or an 
Atomic Safety and Licensing Board, designated by the Commission or by 
the Chairman of the Atomic Safety and Licensing Board Panel, will rule 
on the request and/or petition; and the Secretary or the designated 
Atomic Safety and Licensing Board will issue a notice of a hearing or 
an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
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 [[Page 18640]] 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, 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, Docket No. 50-245, 
MillstoneNuclear Power Station, Unit 1, New London County, 
Connecticut

    Date of application for amendment: March 17, 1995
    Brief description of amendment: The amendment revises Technical 
Specification (TS) Surveillance Requirement 4.7.D.1.c.1 by replacing 
the once per quarter stroke test for containment isolation valves 
(CIVs) with the requirement that the CIVs be tested in accordance with 
the inservice testing program. In addition, there are some editorial 
changes, minor renumbering of subsections, to reflect the TS revisions.
    Date of issuance: March 21, 1995
    Effective date: As of the date of issuance to be implemented 
immediately
    Amendment No.: 81
    Facility Operating License No. DPR-21. 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 March 21, 1995.
    Local Public Document Room location: Learning Resource 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: Phillip F. McKee
    Dated at Rockville, Maryland, this 5th day of April, 1995.
    For the Nuclear Regulatory Commission
Elinor G. Adensam,
Acting Director, Division of Reactor Projects - III/IV, Office of 
Nuclear Reactor Regulation
[Doc. 95-8845 Filed 4-11-95; 8:45 am]
BILLING CODE 7590-01-F