ML17056A201

Discusses Plant Restart Action Plan.Understands That City of Oswego Agreed to Maintain Restart Action Plan in Location Which Allows Easy Access to Plan by Public.Nrc Scheduled Meeting on 890823 to Receive Public Comments on Util Plan

ML17056A201
Person / Time
Site:	Nine Mile Point
Issue date:	07/27/1989
From:	Wiggins J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	Rappaport E OSWEGO COUNTY, NY
References
NUDOCS 8908160113
Download: ML17056A201 (28)

Text

SUL 87 1989 City of Oswego Mayor's Office ATTN:

Mr. Eli Rappaport City Hall

Oswego, New&ark 13126

Subject:

Nine Mile Point Unit 1 Restart Action Plan In December

1987, the Nine Mile Point, Unit 1 Nuclear Power Plant was manually shut down as a result of a technical problem encountered during routine oper-ations.

As a result of this technical

problem, and other technical and pro-grammatic problems identified by itself and by the Nuclear Regulatory Commission (NRC) staff, the Niagara Mohawk Power Corporation elected to main-tain the plant in a

shutdown condition while those problems were addressed and resolved.

The NRC determined that it would require the utility to develop a

formal Plan to address those problems and forward that Plan to the NRC staff for. review and approval.

Subsequently, the utility submitted a document titled the'ine Mile Point, Unit 1 Restart Action Plan, to identify those actions to be completed prior to the restart of the facility.

In the interest of affording local and state officials and members of the public an opportunity to participate in the Plan review process, the NRC has scheduled.-a:meetirig to receive"publi'c comments-on the Utility's Plan.

This meeting is scheduled for August 23,

1989, from 7:00 p.m.

to 9:00 p.m., at the Oswego Middle School,

Oswego, New York'or those interested people who are not able to attend this meeting, written comments may be submitted to me by August 24, 1989, at the following address:

U.S. Nuclear Regulatory Commission, Region I ATTN:

Mr. James. T..Wiggins, Chief Reactor Projects Branch No.

1 Division of Reactor Projects 475 Allendale Road King of Prussia, Pennsylvania 19406 To facilitate public review of the Plan, I understand that you have agreed to maintain copies of the Restart Action Plan in a location which would allow easy access to it by the public.

Therefore, enclosed you will find three (3) copies of the Plan and other related documents.

~i9AQ'J,(iQ], 13 890727 PDR AGCiCK 050002~~i P

PDC OFFICIAL RECORD COPY OSWEGO MAYOR'S OFFICE 0001.0.0 07/27/89

'Nr. Eli Rappaport JUL 4 7:1989 I appreciate your interest and cooperation in this matter.

Sincerely,

Enclosure:

As stated James T. Wiggins, Chief Reactor Projects Branch No. I Division of Reactor Projects cc w/o encl:

W. Davis, Executive Deputy Commissioner, New York State Energy Office bcc w/o encl:

W.

F.

Kane, RI W. A. Cook, SRI Nine Mile Point Units I and 2

K. Abraham, RI.

M. T. Miller, RI M. J.

DiDonato, RI R. A. Capra, NRR RI:DRP Wiggins/mjd 7/~g/89 7/;7 /89 OFFICIAL RECORD COPY OSWEGO MAYOR'S OFFICE 0002

~ 0.0 07/27/89

Pebruary 2 ~ 1989 DISTRIBUTION PDI-1 Rdg CVogan JNeighbors MSlosson MHaughey DLaBarge

~qc)set~file DOCKET NO(S).60-333 31'0-247 50-286 220 0

SUBJECT:

MONTHLY OPERATING ~ORE The following documents concerning our review of the subject facility are transmitted for your information.

Notice of Receipt of Application, dated Draft/Final Environmental'tatement, dated Notice of Availability of Draft/Final Environmental Statement, dated Safety Evaluation Report, or Supplement No.

dated Environmental Assessment and Finding of No Significant Impact, dated Notice of Consideration of Issuance of Facility Operating License or Amendment to Facility Operating License, dated Bi-Weekly Notice; Applications and Amendments to Operating Licenses Involving No Significant Hazards Considerations, dated

[see page(s) j Exemption, dated Construction Permit No.

CPPR-

, Amendment No.

dated Facility Operating License No.',

Amendment No.

dated Order Extending Construction Completion Date, dated

+ Monthly Operating Report for transmitted by letter datedsee a tachment.

Annual/Semi-Annual Report-transmitted by letter dated

Enclosures:

As stated Office of Nuclear Reactor Regulation cc:

aee next page OFFICE)

SURNAME/

DATE/

m PDI-1 HRC I ORM 318 IIOIBDINRCM 0240 OFFICIAL RECORD COPY

October ll, 1988 DOCKET NO(S).

50-220 DISTRIBUTION

..Docket File w'/enclosure PDI-1 Rdg.

CVogan MHaughey JScinto

SUBJECT:

NIAGARA MOHAWK POQER CORPORATION Nine Mile Point Nuclear.Station.

Unit No. l The following documents concerning our review of the subject facility are transmitted for your information.

Notice of Receipt of Applicati on, dated Draft/Final Environmental Statement, dated Notice of Availability of Draft/Final Environmental Statement, dated Safety Evaluation Report, or Supplement No.

dated Environmental Assessment and Finding of No Significant Impact, dated Q Notice of Consideration of Issuance of Facility Operating License or Amendment to Facility Operating License, dated

+ Bi-Meekly Notice; Applications and Amendments to Operating Licenses Involving No d

C id i

.d d~hg gwmftj 4/

Exemption, dated Construction Permit No.

CPPR-

, Amendment No.

dated Facility Operating License No.

, Amendment No.

dated Q Order Extending Construction Completion Date, dated Monthly Operating Report for transmitted by letter dated Annual/Semi-Annual Report-transmitted by letter dated

Enclosures:

As stated Office of Nuclear Reactor Regulation See next page OFFICE/

SURNAME(

PATE)

DT-1 ZPrs ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

w ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

CVo g an 1 0 / 1 1 / 8 8

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

NRC FORM 3IS IIOI80)NRCM0240 OFFICIAL RECORD COPY

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Bfwealdy lfotfce Appffcatfosta anc}'ntencfcnenta to Opcratfno Ucensea'nvofvtno No SfgnNcant Hezarcfis Con slderat)ocis

'. Background Pursuant to Public Law (PK) N'-415.

the Nuclear Regulatory Commission (the Commission) is publishing this regular biweekly notice. PL.87-415 revised section 180 of the Atomic Energy Act of 1954, as amended (the Act). to reqmre 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'ake immediately effective any amendment to an operating. license upon a determination by the Commission that such amendment involves no significant hazards cansideratfon, notwithstanding the pendency before the Commisaioit of a request for a hearing from an@person.

This biweekly notice indudea al) notices ofamendments issued; oc proposed to be issued from September 10, 1988 through September 23, 1988. The last biweekly notice was published an September 21.1988 ($3 FR 36666).

NOTICE OF, CONSIDERATIONOP ISSUANCE OF AMENDMENTTO FACILITYOPERATING UCENSEAND PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATIOM DKrEZMINATION'AND OPPORTUNÃFY FOR HEARING The Conunission has made a pray osed determinatfon that the, fonowing amendment requests involve ne signiTicant hazards consideration. Under the Commission's regulations in10 CFR 5tLq2, thia means that oparatioa of the facilityin accordance with the proposed amendments would not (1) involve a, significant increase in the probability ot consequences ofan accident previously evaluated; or(2) create thepossibifttyof a new or different kind of accfcfent frorrr any accident prevfousfy evahated; or (3) involve a significant reduction fn a.

margin of safety. The bashr for this proposed determinatfoir for each amendment request is shown below.

The Commission is seeking public comments on. this proposed.

determinatioa Any comments received within30 days after the date of publicationof tbis noticewil) be considered in making any final datennfnatiotr. %le Commission wiHrrot notice()y make a Seal detennfnation unless ft receives e request fora heatf~

VPritten commen+nnty be submitted by mail to the Regiihttrry Pablicatfornr Brancir, Division ofFreed om of Information an&Pubffcatfomr Services, Office oFAd'minfstratfon and Resources Management, US. NucfearKeguhtory Commission, Vfashfngton,.DC 20555, and should cfta tFia publication date and page number of this Federa?.Register notice. VAitteircomments may also be dehvered to Room P-216, Phillips BuHdinff, 7920 NorfblkAvenue, Bethesda. Maryland from MO.a.m. to 415 p.m. Copjes ofwritten comments received may he examined at the NRC Public Document Room, the Gelman Building, 2120 L Street, NW..

Washington, DCThe fiifngofrequests for hearing and petitions for leave to intervene is dfscussecf below.

By N'ovember 4. M88, the licensee inay file a request for a hearing with-respect to issuance ofthe amendment to the subject facilityoperating license and any person. whose interest may be affected by this proceeding and who wishes to participate as a party in tha proceeding must filea written petition for leave tn intervan. Requests for a hearing and petitions for leave to intervene shall ba filed in accordance with the Commission's "Rules of Practice for, Domestic. Licensing Proceedings" in M CFR Part 2. Ifa request for a hearing or petition for leave to intervene is Bled by the above date. the Commission or an~Atomic Safety and Licensing BoarcL designated by the Cammfssioa or by the Chauman.

of tha AtomicSafety and Licensing Board.Pawl wil)rule on the request and/or, petition and the Secretary or the designated Atomic Safety and:Licensing Board wfILissue ct notice of hearing or an appropriate order.

As required by 19 CFB M14, a petition for leave ta intervene shall sel forthwithparticuhirity the interest of the petitioner in the proceeding, end how that interest may be. affected by the results of the proceeding. The petition should speciTicafiy explain the reasons why intervention'should be permitted partfcuhLrroferaaca to the fo)fowfngfactor/r. (1) the nature of the petihonar's right eader tbe Aet to be made a party to tbeproceedfng,"(2) the nature and extent of the petitioner's property, financiQ orether fnterest ia the proceeding; acid/~ the possible effect of any order whk% may be entered in the proceeding on the petitioner's interest. The petilfon should also identify the speal'spect(s) of tha

Feder'sl'e8(ster o)'3,.No; -193 g Wechesd'ay;-Ootober 5,"4888 /"Nbtfc'es subject matter of the. proceeding as to'hich petitioner wishes to Intervene.

Any person who has filed a petition for leave to intervene or who has been admitted as a party mayamend the petition without requesting leave of the Board up to fifteen (15) days prior to the'irst prehearing conference scheduled In the proceeding, but such an amended petition must satisfy the specificity requirements described above.

Not later than fifteen (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. and the bases for each contention set forth with reasonable specificity. Contentions shall be limited to matters within the scope of the amendment under consideration. A petitioner who fails to file such a supplement which satisfies these requirements with respect to at least one contention willnot 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 fullyin the conduct of the hearing. including the opportunity to

= present evidence and cross-examine witnesses.

Ifa hearing is requested, the Commission willmake a final determination on the issue ofno significant hazards consideration. The final determination willserve to decide when the hearing is held.

Ifthe 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, Ifthe final determination is that the amendment involves a significant hazards consideration, any hearing held would take place before the issuance of any amendment.

Normally, the Commission willnot issue the amendment until the expiration of the 30-day notice period.

However, should circumstances change during the notice period such that fa0ure 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 signifiicant hazards consideration. The final determination willconsider all public and State comments received before action Is taken. Should th' Commission take this action; Itwill' publish a notice of issuance and provide for opportunity for a hearing after issuance. The Commission expect~'that the need to take this action willoccur very infrequently.

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 2055S, Attention:

Docketing and Service 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 ten (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) 325-6000 (in Missouri 1-(800) 342-6700). The Western Union operator should be given Datagram Identification Number 3737 and the followingmessage addressed to (Pleat Director)i 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 hearing willnot be entertained absent a determination by the Commission. the presiding officer or the presiding 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)-fv) 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 facilityinvolved.

Commonwealth Edison Company, Docket No. 50-237, Dresden Nuclear Power Station, Unit No. 2, Grundy County, Illinois Date ofoppiicotian foramendment requestt August 2S, 1988 Description ofamendment request:

Commonwealth Edison Company (CECo) has proposed changes to the Dresden Unit 2 Technical Specifications to facilitate future reload licensing reviews per 10 CFR 50.59. These proposed changes are as follows: (1)

Deletion of the license condition requiring a safety evaluation for coastdown operation with off-normal feedwa ter temperature from Section 3.E of the license; (2) Revision of the'inimum Critical Power Ratio (MCPR) operating limitto a conservative value likely to bound cycle specific results for the neer term; (3) Revision of the Single Loop Opeiation (SLO) MCPR adder to 0.01 (from 0.03) and a revision in the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) reduction factor for SLO to 0.91 (from 0.70): (4)

Incorporation ofTransient Linear Heat Generation Rate (TLHGR) limits; (5)

Revisions ofreduced flow MCPR limits:

(8) and Revision of the relief valve Technical Specification to require action only after two relief valves are found to be inoperable, provided MAPLHGR reduction factors are implemented.

In addition, proposed administrative Technical Specification changes have been provided which include: changing references to Exxon Nuclear Company (ENC) to advanced Nuclear Fuels Corporation (ANF), except in titles of earlier documents and definitions of.

nuclear limits. and defining Transient Linear Heat Generation Rate (TLHGR),

Steady State LHGR (SLHGR), LHGR.

Fuel Design LimitingRatio for Centerline melt (FDLRC), and Fuel Design Limiting Ratio for Exxon Fuel (FDLRX).

The amendment application ofAugust

25. 1988 is supported by the following analyses which were submitted: ANF Document, XN-NF-84-49, "Analysis of Dresden Units 2 and 3 Operation with One Relief Valve Out-of-Service", dated September 1984; ANF-87-111, "LOCA-ECCS Analysis for Dresden Units During Single Loop Operation with ANFFuel",

dated September 1987: ANF-88-79(P).

"Dresden ReportMechanical, Thermal, and Neutronic Design for ANF9x9 Fuel Assemblies", dated May 1988: ANF 69, "Extended Operating Domain/

Equipment Out-of-Service Analysis for Dresden Units 2 and 3". dated July 1988:

and GE Letter, REP: 88-181. R. E Parr to R. A. Roehl. "Correction to Dresden 2 Cycle 12 Alternate Water Chemistry LTA's MAPLHGR Curve", July 26. 1988.

These analyses are similar or identical to the analyses that were previously submitted by CECo for the Dresden 3 Cycle 11 reload and approved by the staff on June 20, 1988.

Basis forproposed no significant hazards consideration determinatiom The Commission has provided standards for determining whether a significarit hazards consideration exists as stated in 10 CFR 50.92(c). A proposed amendment to an operating license for a facilityinvolves no significant hazards

Federal RegtOs

/ Vol. 55, No. 195./ Wednesday. OctO 5. 1988 / Notices SM67 considerations ifoperation of the facility in accordance with the proposed amendm'ent would not; (1) involve a significartt increase in the probability or consequences of an accident previously evaluated; (2) create the possibility of a new or different kind ofaccident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

The licensee addressed the above three standards in the amendment application as follows:

(a) Involve a significant increase in the probability or consequences of an accident previously evaluated because:

Relative to Item 1:

ANF has performed analyses with NRC approved methodologies to ensure that transients occurring under coastdown conditions with off-normal feedwater temperature are bounded by transients at rated conditions.

Relative to Item 2:

The incorporation of the proposed MCPR operating limits noted above is provided to establish limits on reactor operation which ensure that the core is operated within the assumptions and initial conditions of the transient analyses. Operation within these limits willensure that the consequences of a transient or accident remain within the results of the analyses. The probability of an accident is not affected by this change because no physical systems or equipment which could initiate an accident are affected and the MCPR safety limitcontinues to be protected.

Relative to Item 3:

The incorporation of the proposed MCPR and MAPLHGRlimits during single loop operation establishes limits on reactor operation to ensure thermal-mechanical integrity of the fuel and cladding. Neither the consequences nor the probability of an accident is affected by this change because the design basis transients and accidents were considered when establishing these operating limits.

Relative to Item 4:

The incorporation of the proposed TLHGR limits establishes limits on reactor operation to ensure thermal-mechanical integrity of the fuel under transient overpower conditions, consistent with the fuel vendor's design criteria and the surveillance method already performed in the onsite core monitoring computer software.

Consequences of previously evaluated events are therefore not affected, The probability of an accident is not affected by this change because no physical systems or equipment which could initiate an accident are affected and the cladding integrity willbe maintained during overpower events.

Relative to Item 5:

The incorporation of the proposed.

reduced flow MCPR limits establishes limits on reactor operation to ensure that thermal limits willnot be violated during transients initiated during off-rated core flows. Therefore, consequences of postulated events are unaffected. The probabiflty of an accident is not affected by this change because no physical systems or equipment which could initiate an accident are affected and the MCPR safety limitwillbe protected during overpower events initiated at off-rated core flows.

Relative to Item 8:

ANF has performed analyses with NRC approved methodologies to ensure that reactor thermal limits are not violated during limitingtransients with one relief valve outwf-service. Event consequences are therefore not affected by this change. The probability of an accident is not affected by this change because no physical systems or equipment which could initiate an accident are significantly affected.

Relative to Items 7 and 8:

These changes are administrative in nature and have no impact on any systems or limits on reactor operation.

(b) Create the possibility of a new or

'ifferent ldnd of accident from any.

accident previously evaluated because:

Relative to Item 1:

ANF has determined that transients occurring at off-rated feedwater heating during coastdown ar'e bounded by those initiated at rated, fullpower conditions.

Furthermore. there is no impact or physical modifications to systems or components whose failure could Initiate a new or different kind of accident, Relative to Items 2. 3, 4. and 5:

The proposed MCPR, MAPLHGR, and LHGR limits represent limits on core power distribution which do not directly affect the operation or function of any system or component. As a result, there is no impact on or addition of any systems or equipment whose failure could initiate a new or different kind of accident.

Relative to Item 8:

Operation is allowed with one relief valve out-of-service (RVOOS) provided appropriate MAPLHGRreductions are implemented. This change in no way impacts the function of the remaining operable valves or other equipment and since the appropriate requirements to test HPCI are included, this change does not create a new or different ldnd of accident.

Relative to Items 7 and 8:

These changes are administrative in nature and have no impact on or modification t'o any system or equipment whose failure could initiate

'n accident.

(c) Involve a significant reduction in the margin. ofsafety because:

Relative to Item 1:

The analysis supporting this change shows that transients during coastdown with off-normal feedwater temperature are bounded by transients at rated conditions, therefore no reduction in the margin of safety occurs.

Relative to Items 2, 3. 4. and 5:

These changes have been analyzed to demonstrate that the consequences of transients or accidents are not increased, using the specified restrictions, beyond those previously evaluated and accepted at Dresden. The analyses show that the MCPR safety limit,fuel thermal-mechanical limits, and reactor pressure limits are not violated during postulated transients.

Relative to Item 6:

Previous analysis supporting this change has shown that the point of minimum MCPR occurs before any relief valves open, indicating the assumption of one relief valve out-of-service willnot reduce the margin to safety for anticipated abnormal operating transients. For LOCA, analysis has shown that with the speciflc MAPLHGR restrictions, all criteria of 10 CFR 50.46 are satisfied for the limitingsmall break.

Large breaks are unaffected.

Relative to Items 7 and 8:

These changes are admini'strative in nature, either deleting information that is no longer applicable or providing clarification to current specifications.

The staff has reviewed the licensee's no significant hazards analyses given above. Based on this review, the staff proposes to determine that the proposed amendments meet the three 10 CFR 50.92(c) standards and do not involve a significant hazards consideration.

Local Public Document Room location: Morris Public Library. 604 Liberty Street, Morris, Illinois60450, Attorneyforlicensee: Michael I Miller,Esquire; Sidley and Austin, One First National Plaza, Chicago, Illinois 60603.

NRC Project Director: Daniel R.

Muller Consumers Power Company, Docket No.

50-155, Big Rock Point Plant, Charlevoix County, Michigan Dale ofamendmenl request:

September 8, 1988.

Description ofamendment request/

This amendment would delete Figure 6.2-1. "Offsite Organization," and 8.2-2, "Plant Organization." from the Big Rock Point Plant Technical Specifications and would indicate where those figures will

Federal Registei Vol. 53, No. 193 / Wednesday, October a, X95LJ.Noticea....,...

hereafter be maintamed. It would also augment the text of Section 8, "Administrative Controls." ta incorporate responsibilities of the key positions affecting safety. change the title "Plant Superintendent" to "Plant Manager," and make such other minor changes as necessary to ensure that the requirements for offsite and onsite organizations are adequately described.

Basis forproposed no significant hazards cansiderati an determi nations The Commission has provided standards in 10 CFR 50.92(c) for determining whether a signiTicant hazards consideration exists. A proposed amendment to an operating license for a facilityinvolves no significant hazards consideration if operation of the facilityin accordance with the proposed amendment would not: (1) involve a significant increase in the probability or consequences ofan accideiit previously evaluated, (2) create the possibility of a new or different kind of accident fram any accident previously.

evaluated, ar (3) involve a signiTicant reduction in a margin of safety. The Consumers Power Company (CPC) reviewed the proposed change and determined, and the Commission's staff agrees, that.

(1) The proposed amendment does not involve a significant increase-in the probability or consequences ofan accident previously evaluated because deletion of the organization charts from the Technical Specifications does not affect plant operation. The revised Technical Specifications require that "Lines ofauthority, responsibility and comnumication... established and defmed for the highest management levels through intermediate levels to and including operating organization positions... shall be documented, updated and reported to the NRC..."

(2) The proposed amendment does not create the possibility of a new or different kind of accident than previously evaluated because the proposed change is administrative in nature and no physical alterations of plant configuration or changes to setpoints or operating parameters are proposed.

(3) The proposed amendment does not involve a significant reduction in a margin of safety because CPC. through its quality assurance programs, its commitment to maintain only qualified personnel in positions of responsibility, and other required controls, assures that safety functions willbe performed at a high level of competence. Therefore, removal ofthe organization charts from the Technical SpeciTications willnot affect the margin of safety.

Accordingly, the Commission proposes to determine that this change does not involve a significant hazards consideration.

Local Public'Dacament Raom location: North Central Michigan College, 151S Howard Street. Petoskey, Michigan 49770.

Attorrieyfar licensee: Judd L Bacon, Esquire, Consumers Power Company.

212 West Michigan Avenue, Jackson.

Michigan 49201.

NRC Project Director: Martin J.

Virgiflio.

Duquesne Light Company, Docket No.

50-412, Beaver Valley Power Station, Unit No. 2, Shippingport, Pennsylvania Date ofamendment requests August 11, 1988 Description ofamendment request:

The proposed amendment would revise the supplemental leak collection and release system (SLCRS) fi!ter trains surveillance requirements. Technical SpeciTica tion 4.7.8.1.b currently requires an in-place leak test be performed on the HEPA filterand charcoal adsorbers, and an iodine removal efficiency test be performed on the adsorber stage at least once per 18 months. Both the leak tests and the iodine removal efficiency test are also required followingpainting, fire or chemical release in any area commumcating with the SLCRS. The proposed change wonMmodify the filter testing requirements such that following painting. Bre or chemical release in these areas, only the iodine removal efficiency test would be required. The licensee stated that because of the unique design of the Beaver Valley Unit 2 SLCRS, elimination of the HEPA filter test and charcoal adsorber test can be justified.

This proposed amendment would also revise the SLCRS flowrate from 59,000 CFM %10% to 57,000@PM %10%. This change is a result ofremoving the main steam and feedwater valve area from SLCRS coverage, and also reflects the actual system flowrates obtained after final system balancing. There is no piping in the main steam and feedwirter valve area which couR contain post.

LOCA fluids. Therefore the capability of the SLCRS to collect radioactive effluents from ESF systems operating outside the containment followingany postulated LOCAs wiflnot be affected.

Basis forproposed no significant hazards consideration determinationt The Commission has provided standards for determining whether a significant hazards consideration exists (10 CFR 50.92(c)). A proposed amendment t" an operating license for a facilityinvolves no significarit hazards consideration ifoperation of the facility in accordance with the proposed amendment wauld not (1) involve a significant increase in the probality or consequences of an accident previously evaluated; (2) create the iiossibllityof a new or differerit kind of accident from any accident previously evatuateth or (3) involve a significant reduction in a margin of safety.

The proposed changes are not made as a result of, nor would they lead to any SLCRS design changeL VVhen approved by the staff, only unnecessary surveillance requirements would be eliminated. The SLCRS willcontinue to perform as stated in the licensee's Final Safety Analysis Report. Therefore, the answer to both questions (1) Ec (2) is negative. The amended requirements willcontinue to ensure the operability of the SLCRS; there would be no relaxation afpreviously used safety margins.

Therefore, the answer to question (3) is also negative.

On such basis, the staff proposes to determine that the requested amendment involves no significant hazards consideration.

Local Public Document Room location: B.F. Jones Memorial Library.

663 Franklm Avenue, Aliquippa.

Pamsylvania 15001.

Attorneyforlicensee: Gerald Charnoff, Esquire, Jay E. Silberg.

Esquire. Shaw, Pittman, Potts 8 Trowbridge, 2300M Street, NW.,

Washington. DC 20037.

NRC Project Director, John F. Stolz Duqaesne LightCompany, Docket No.

50-412, Boaver Valley Power Station, Unit No. 2, Shippingport, Pennsylvania Date afamendment request: August 30, 1988 Description ofamendmenl request:

The proposed amendment covers a number of pages in the Technical Speciflcations addressing allowable enrichments and canfigurations far fuel stored in the spent fuel starage pool The proposed changes are:

1. Section 33414 and an accompanying Table 3.9-1 would be added to specify allowable enrichment and configurations for stored%el

2. Basis Section 3/4.9.14 would be added to provide the bases for the above specifications.

3. Section 5.3.1 would be amended to specify a higher enrichment of 4.85 weight percent U-235 (currently 3.3 weight percent), and

4. Section 5.8.1 would be revised to reference appropriate sections in the FSAB where the spent fuel pool criticalityanalysis can be found.

The proposed new speciTicationa, with associated guidance incorporated into

Fsdesal Regl/

VoL 88, No. 198 / Wednesday, Oct+ sr 1988 / No8ces existing administrative. controls would

'ermit storage offuel with up to 4.85 weight percent U-235. The pool would be separated into two regions. Spent fuel pool region 1 would provide for storage of fuel with enrichments up to 4.85 weight percent U-235 in an administratively controlled 3-of-1 cell array. Region 2 would provide for storage offuel assemblies with the burnup-dependent enrichment limitations provided in Table 3.9.1. Also, the boron concentration in the spent fuel pool would be specified to be maintained at greater then or equal to 1050 ppm when moving fuel in the spent fuel pool. Suberftlca]ity would be maintained by limitingfuel assembly interaction and maintaining the minimum boron concentration.

Basis forproposed no significont hazards consideration determinotion:

The Commission has provided standards for determining whether a significant hazards consideration exists (10 CFR 50.92(c)). A proposed amendment to an operating license for a facilityinvolves no significant hazards consideration ifoperation of the facf]lty in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated: (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a signiifiicant reduction in a margin of safety.

There is no change in fuel pool hardware, but the assodated Updated Final Safety Analysis Report willbe changed to Include analyses to demonstrate that the fuel pool and stored fuel willcomply with unchanged performance objectives and limitations (e.g.. criticalityand heat dissipation).

The criticalityanalysis acceptance criteria (K,/F less than 0.95) is consistent with that stated in the PSALM The segregation of the spent fuel pool Into regions 1 and 2 and appropriate administrative constraints ensure that analysis assumptions are valid and that performance criteria would be met when fuel is not being moved. In addition to the administrative constraints available to maintain appropriate fuel storage configurations. the minimum boron concmitration willensure that criticality willnot be achieved even iffuel assemblies were not stored in the specified checkerboard arrays. Fuel assembly decay heat production is a functfon of core power level, and since the core power level would remain unchanged, the decay heat load on the spent fuel pool cooling system would not be affected by the proposed enrichment-limits.

The radiological consequences'of the

~

fuel handling accident are dependent, among other factors. upon power level of the reactor. There is no power level change associated with the proposed amendment and since all other factors would not be changed by this amendment, the consequences of the fuel handling accident would not be changed.

No hardware modification is involved and the changes to existing administrative controls involve only prescription of the loading patterns to accominodate a greater variety of fuel assembly enrichments without change in performance. There is thus no increase in the probability of the fuel handling accident previously analyzed in the FSAR. and there is no possibility of a new type of accident different from any previously evaluated. Furthermore, there is no change In any acceptance criterion as stated above; therefore, there is no reduction of a safety margin.

Accordingly, the staff has made a proposed determination that the requested amendment does not involve a signfficant hazards consideration.

Local Public Document Room locotian: B. F. Jones Memorial Library, 883 Franklin Avenue, Aliquippa, Pennsylvania 15001.

Attorneyforlicensee: Gerald Charnoff, Esquire, Jay F Silberg, Esquire, Shaw, Pittman, Potts 59 Trowbridge, 2300 N Street. NW.,

Washington, DC 20037.

NRC Prof ect Director: John F. Stolz Florida Power and Light Company, et aL, Docket No. 50-389, St. Ludo Plant, Unit No. 2, St. Ludo County, Florida Date ofomendment request:

September 1. 1988 Description ofamendment request:

The proposed amendment would make changes to the Technical Specifications associated with the boric acid makeup (BAMU)system. Spedfica))y, the required boron concentratfon requirements would be reduced, the borated water volume would be increased. and the requirement to heat trace the BAMUsystem would be deleted.

Basis forproposed no significant hazards consideration determination:

The Commission has provided standards for determining whether a significant hazards consideration exists as stated in (10 CFR 508I2(c)). A proposed amendment to an operating license for a facilityinvolves no significant hazards considerations ff operation of the facilityin 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 ofaccident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

The licensee addressed the above three standards in the amendment applfcation. In regard to the first standard, the licensee provided the following analysis:

1. Involve a sfgnfficant increase In the probability or consequences of an accident previously evaluated.,

The operatfon of the facility In accordance with the proposed changes does not Involve a sffpiificanthcresse in the probability or consequences of any accident previously evaluated. Deleting the requirement for a heat tradng drcuft by redudng the boron concentration In the [boric acid makeup tanks) ((BAMTs))Is accounted for by Increasing the volume of boric acid solution that must be contained In the tenks and by also credftfng borated weier froin the

[refueifng water tank] ((RWT)).Since the components [or their function) necessary to perform a safe shut down have not been changed or modffied, this change does not sfffnificsntlyIncrease the probability or consequences of any acddent previously evaluated. In addftlon. administrative controls on the boric acid makeup tank temperature and boron concentration ensure that the lack of heat tracing does not result in precipltatfon of the boron.

The reduction In boric add concentratfon in the boric acid makeup tanke hae been evaluated to determine the effect of this reduction on containment sump.pH and boric acid concentration. The existing post LOCA

[c]ontafnment [a]ump [f]nventory calculation was recalculated to reflect the new operating parameters as a result of the reduction In boric acid concentration fn the [b]oric[a]dd

[m]akeup tanks. The results of the caicuiation esiabffsh that the Post LOCA long term contafnment sump and spray chemfstry shall have new boundfng values for boric acid concentration and pH. The Equipment Qualificatfon Documentation Packages were reviewed to determine ifthe new boric sdd concentratfon and pH ranges are bounded by the currently specified ranges for

[e]nvfronmental [q]ueflficaifonduring a LOCA. Thc determination Is that the equipment In the containment can be quafffied for the bounding values of the boric acid concentrations and pH valuee.

An evaluation was performed to determine the effect of the new pH range en mechanical systems and components due to corrosion. By mafntalnfng the pH of the long tenn Contefnment Sump and Spray System io between 7.0 and eAh the evolution of iodine and the effect of chforlde and caustic stress ere minimized.

Credit ls not taken for boron addition to the reacto'r coolant system from the boric acid makeup tanks for the purpose of reactfvlty control in the accidents analyzed In Chapter 15 of the pisnt's Final Safety Analysis Report. '

S9178 Fedarai RegisiiiOVOI. dii, No. 1!}ii /'Wedni.Sday, OCiOiiii~,

1988$

NOGCeii Reaponsetosuch events assteam line break overcooling, boson dUuhon. etc., willnot be affected by a reduction in the BAMT concentraUon. In particular. the action statements associated with Technical Specification 3.1.12 require that boration be commenced at greater than 4D gaUons per minute using a solution of at least 1y20 ppin boron in the event that shutdown margin is lost. As noted before the BAMTboron concentraUon after it Is reduced willbe in cxress of1720 ppm.

In connection with the second standard. the licensee stated:

2. Create the possibility of anew or different kind of accident from any accident previously evaluated.

The operation of the facilityin accordance with the proposed changes does not create the possibigty of a new or different kind of accident from any accident previously evaluated. The mason for requiring a heat tracing circuit nas to ensure that the dissolved boric acid was in solution and hence. available for injection into the Reactor Coolant System (RCS) to adjust core reactivity throughout core life. By lowering the boroncancentration to a maximum of 3.5 weight percent, chemical analyses have shown there is no possibility of the boron precipitating oat of solution as long as the temperatiire of the boric acid remains above 50'; thus there is no longer a need for heat tracing. Since tbe boron willbe in solution when the BAMTflowpaths are credited for reactivity control during the safe shutdown scenario. heat tracing is no longer required to inaintain tbe (b)oric )a]cid (rn)akeup system operable. In conclusion, this change does not c.eats the possibility of a new or different kind of accident from those previously evaiuatetL With regard ta the third standard, the licensee provided the following rationale:

3. Involve a slgnificant reduction in a margin of safely.

The operation of the facilityin accordance w'ith the proposed Technical Specification changes does not involve a significant reduction In the margin of safety. The intent of these Techmcal Specifications is to ensure that there are two redundant Ilowpaths from the borated water sources (BAMTsand RWT) to the reactor coolant system to allow control ofcore reactivity throughout core life.

This requires that sufficient quantities of boron be stored in the BAhITs and that this borated water can be delivered to the RCS in the event of a single active failure oF a system component or a seismic event. Reducing the maximum boric acid concentration to less than 3.5 weight percent has been compensated for by increasing the required minimum volumes of borated water. In addition. reducing the maximum baron concentration allows a deietian of ibe requirement io hest trace the Ib)eric {a)cid (m)akeup system since chemical analyses have shown that ~ 3$ weight solution of boric acid willremain in solution at temperatures above 50'. Administrative rontrobion tbe boric acid niakeap tank temperature and boron concentration ensure that a lack of heat tracing does not result in precipitation of the boron. In oonciusion. the reduction oF boric acid concentration and <he deletion of heat tracing In the (b)oric (a)cid (ra)akeup system does not.cause a significant reduction in the margin of safety for this plant.

The staff has reviewed the licensee's no aigniTicant hazards canaideratiau determinatiaa analysis. Based upon this review. the staff believes that tho licensee has met the three standards.

The licensee proposes increaaea,ta the borated water volume contained in the BAMTs to offset the boron concentration reduction. Since the boron concentration is signiTicantly reduced, there appears ta be na need forheat tracing.

Based upon the above discussion, the staff proposes ta determine that the propased changes dn stat involve a signiTicant hazards consideration.

Local Public Daaumeut Room locatianr &dian River Junior Collage Library. 3209 Virginia Avenue. Fart Pierce, Florida 33450 Attorneyforlicensees Hara)d F. Rais, Esquire, Newman and Haltzinger, 1615 L Street, NW.. Wastungtan, DC20038 NRCProject Direatar: Harbor t N.

Berkaw Gulf States UIIIIUas Company, Docket Na. 50-458, R)vov Baud Station, Unit 1, West Feliciana Parish, Loaudana Dale ofamendment neqaect: August 5, 1988 Description ofamendment requesL The amendment request would utadlfy (1) License Condition? C(13), and (2)

Technical SpeciTicatlan Table 3.3.6-2, Item 1.b, High Power Seipaint ta allow continued operation af the facilitywith up to 100' reducticm fmm the rated feedwater temperature ef428' during the normal fuel cycle. Pl'armed aperaGan with partial faedwater heating for the purpose of extending the fuel cyc)e would continue ta be prohibited. Liceiise Condition 2l:(13) would be modtTiad to read, "The facihty shaII not be operated with psr tlal feedwstor heating beyond the end of the siarmal fuel cycle without prior approva) of the staK During the normal fuel cycle, the facilityshall not be operated with a feadwaterhaating capacity which would result in a rated thermal power feedwater teinperatura less than 320' without prior approval of the staff." Technical Spec)Qcatian Table 3.3.6-2. Stern LkHigh Power Setpaint, would he modified as fallaws:

(1) the trip setpaiut id'auld be &iMtgad10

,less than or equal ta 67'%f rated thermal power from the current value of 63.5~3%, and (2) the allowable value would be changed ta less than orequal to 682% ofrated thermal power from the current value af 825~~4%.

Basis forprapased no significant hazards consideration determinatian:

The Cammlsslan has provided

'tandards for determining whether a signiflctttithazards consideratlan exists as s tatmiin 10 CFR 50.92(c)

~ A proposed amendment ta an aperating license for a facilityInvabres na significant hazards consideratian ifoperation. ofthe facility in accordance with the proposed amendmentmau)d not: (1) ittvalve a significant increase in the probability or consequences of an accident previously eva)uated: or (2) create the possibility of a new or different kind of accident from any accident previous)y evaluated: or (3) involve a significant reduction in a margm ofsafety. The hoensee provided an analysis that addressed the above three s4asdaads in the ameitdment application.

1. No significant IncnMNe in the probability or the consequences of an accident previously evaluated results Rom this request because:

... all (Updated Safety Analysis Report)

USAR Chap1er15 co~de transients were examined forffeedwater heater(a) out of service) FWHOS operation This mode of operation reoistts in decreaseiI headwater tenqiecatme aad bicressed sobcooiing in the core downcoainr region and ai tbe core inlet.

As shown bdoes, the effects of this do siot Increase tbe probability ofany previously evaluated accictants or transients.

1brae IimiUag transients were reevaluated in detaiL They are:

(1) Generator Load Rejection with Bypass Failure (LRBPFI (2) Feedwater Flow Controller Failure.

Maximum Demami (FWCF)

(3) Loss of 100' Feedwater Heating (LAVH)

The results of the evaluations for transients (1) & (2) demonstrate that these delta'(oriUcal power ratios) CPRs are below the lbniting delta CPR of0.11 documented in Section 10 tt 11 of the reload license subinitiel for RBS Cycle 2...'Iberefore. tbe consequences of these events are bounded by the current Technical SpeciTication limits with respect to LRBPF and FWCF events.

The (River Bend StaUon) RBS plant specific analysis For tho1txr F loss offeedwater heating iransient (transient (3} above) for FWHOS operation is adequately bounded by the 420' aoauai feedwaier temperature delta CPR results of0.11...

The )rod withdrawal error) RWE transient analyses were also reevaluated...

the results of this evaluation indicate that the resulting delta CPR Is unchanged from 0.11 Since tbe tcsuIUng delta CPRs for the events analyzed above remain bounded by the limiting4e)ta CPR of011.... the operating limithlCPR (OLhtCPR) does not need to be changed os a icsidt ofa RWR during AVHOSoperation. AddiUonaily. the off-rated poweMependent MCPR, limits are not affected by FWHOSopemUon and remain bounded by ihe current BWE off-rated power dependent MCPR, limits... the off rated fiow-dependent MCPRi limits for

Fedeeet Regis/

VoL 33. No. 193 / Wednesday, Octo, 1988 / Notices FWHO'S operation are bounded by the current sg ICPRF limits.

The consequences of anticipated transient without scram (ATWS) and reactor vessel overpressurization transients are less severe under the initial conditions of partial feedwater heating than that of normal feedwater heating. With reduced feedwater temperature at rated thermal power, the initial steaming rate is less. which would yield less severe results during an ATWS event. Lower initial operating pressure and lower steam flowrate during FWHOS, operation yield lower peak vessel pressure for the most limitingmain steam line Isolation valve closure event.

An evaluation of the impact of FWHOS operation on the RBS LOCA analysis was also performed. The results of this evaluation show that the resulting peak cladding temperature would be lower than the 2144' value reported in USAR Chapter 6 and below the 2200' liinitidentified in 10 CFR 50.46.

Acoustic and flow-induced loads on reactor internals created during a LOCA with FWHOS operation were evaluated... While these loads would increase slightly, the results of this evaluation concluded that there Is adequate conservatism in the evaluation and significant design margin remains available to account for these loads during FWHOS operation.

The impact of FWHOS operation on the containment LOCA response was also evaluated. Both the main steamline break and recirculation line break cases were reanalyzed over the FWHOS operation power/flow region. The peak drywell and wetwell pressure and temperature, pool swell. condensation oscillation and chugging loads were evaluated. The peak drywell-to-wetwefl differential pressure during the FWHOS operation occurred under recirculation line break at the maximum vessel subcooling condition on the power/

flow map. This peak differential pressure increased by 1.02 psi. However. the resulting differential pressuie is stillbelow the design differential pressure of 25 psid presented in USAR Table 6.2-1. Also, the pool swefl.

condensation oscillation, and chugging loads evaluated at the worst power/flow condition during the FWHOS operation vary slightly over the peak values presented in USAR Section 8. The analysis concluded that this variation is insignificant and there is adequate design margin to account for these loads during FWHOS operation.

A study was performed to assess the impact of FWHOS operation on the annulus pressurization (AP) loads forRiver Bend Station. The feedwater line break case results In the greatest forces upon the reactor pressure vessel and the greatest pressure differentials across the biological shield walL The break low for this case with FWHOS operation was determined to be less than that presented In the USAR during the Inventory depletion period when the peak AP loads occur. Therefore, the normal operation AP loads calculated in the RBS USAR bound those expected to result under FWHOS.

An evaluation of the effect of FWHOS operation on the feedwater nozzle at RBS was also performed. Assuming 80% capacity factor with continuous FWHOS operation, the fatigue usage factor for the feedwater nozzle would increase by 00214 over 40 years of continuous FWHOS operation. However, the fatigue usage factor would stiU be less than 0.6, which is below the limitof 1 L A standard stress analysis was performed on the feedwater system piping up to the first feedwater guide lug outside the containment for a bounding feedwater temperature, of 250'.

Results of this study show that with FWHOS operations, the feedwater piping fatigue usage factor is less than that at rated conditions due to a lower temperature gradient through the piping wall.

An evaluation was performed to examine the impact of FWHOS operation on the feedwater sparger for RBS. A case was analyzed to determine the number of days of FWHOS operation allowable per year (for 40 years) without exceeding the feedwater sparger fatigue usage factor Emit of 1.0. The results show that the 40 year average number of days allowable during an operaUng year for FWHOS operation is 256 days for a rated feedwater temperature of 370' and 61 days for a rated feedwater temperature of320'.

Administrative controls to ensure that the number of days and the magnitude of temperature reduction during FWHOS operation is tracked wlflensure that FWHOS operation cannot increase the probability or consequences of any accident previously evaluated.

... with regard to reactor core therma)-

hydraulic stability, FWHOS operation is

~

bounded by the fuel integrity analyses described in... "Compliance of the Genera)

Electric BoilingWater Reactor Fuel Designs to Stability IJcensing Criteria," NEDE-22272-P-1. October 1984. Therefore, the generic operator recommendaUons on thermal-hydraulic stability are still appflcable and adequately address FWHOS operation.

Impact of FWHOS operation on the (turbine stop valve) TSV position and (turbine control valve) TCV fast closure reactor scram bypass setpolnts and the (end of cycle recirculation pump trip] EOC RFf bypass setpoint and the (rod control and information system) RCIS high power and low power setpoints was also evaluated. The required upper bound for bypass of the TSV position and TCV fast closure reactor scrams and EOC RPT is 40% of rated thermal power.

Below 40% rated thermal power. high neutron flux. vessel pressure. and other normal scram functions are sufflcient to provide margin to the safety limits (even with TSV or TCV closures) as identified in USAR Section 15~.2.32 Therefore, below 40% rated thermal power, the TSV and TCV scrams and EOC RPT functions are bypassed.

Turbine first.stage pressure (TFSP) ls the parameter used to activate these reactor scram and EOC RPI'ypasses below 40%

rated thermal power. Under operation with reduced feedwater temperature. the relationship between vessel steam flow(and therefore TFSP) and core thermal power changes. Less steam flow Is generated at the same thermal power and the TFSP Is reduced.

Therefore, the effect of reduced feedwater temperature is to raise the thermal power level forwhich the EOC RPT and scram bypass functions are seL Conservatism in the current RBS Technical SpecificaUon scram bypass TFSP nominal setpoint was assessed by comparing it to the RBS startup test data for 'TFSP vs. Reactor Power." The current Technical Specification setpoint is conservative in the scram bypass power level by approximately 8% for feedwater temperature operation at 420 F and by approximately 4% for FWHOS operation at 320' when coinpared to the setpoint actually required. Therefore, the conservatism In the current Technical Specification setpoint adequately accounts for FWHOS operation.

The proposed change in RBS Technical Specification Table 3.3.8-2. Item 1.b, High Power Setpoint, restricts plant operation to conditions assumed in the RWE analysis and is consistent with the upper range of the allowable value currenlly specified. The proposed change is also consistent with the Standard Technical Specifications and the Technical Specifications ofother licensed BWR/6 plants.

Based upon these considerations. It is concluded that operation with FWHOS and the proposed change to the high power setpoint Technical Specification do not increase the probability or consequences of any accidents previously evaluated.

2. This request would not create the possibility of a new or different kind of accident from any accident previously evaluated because:

FWHOS operation results in decreased feedwater temperature and increased subcooling in the core downcomer region and at the core inlet. As shown in Item 1 above, the impact of FWHOS operation has been found to be adequately bounded by the current analysis provided In the River Bend Station USAR with the exception of the feedwater sparger fatigue usage'factor. The number of days of FWHOS operation must be limited to ensure that the feedwater sparger fatigue usage factor does not exceed 1'dministrative controls to ensure that the number of days and magnitude of temperature reduction during FWHOS operation is tracked willensure that FWHOS operation cannot create the possibility ofa new or different kind ofaccident from any previously evaluated. Addittonafly, FWHOS operation does not involve any hardware changes and is wefl within the capability of existing equipment. Hence, no new failure modes are introduced.

The proposed change in RBS Technical SpecificaUon Table 3DM2, Item 14, High Power Setpoint, restricts operation to conditions assumed in the RWE'analysis and Is consistent with the upper range currently specified. The proposed change is also consistent with the Standard Technical SpedflcaUons and the Technical

. Specifications of other Ucensed BWR/8 plants. Therefore, thh mode of operation does not create the possibility of a new or different ldnd of accident from any previously evaluated.

3. This request would not Involve a significant reduction in the margin of safety because:

As stated in the response to item 1 above, the results of the 320' feedwater temperature FWHOS operation case are bounded by the results of the analyses

39172 yedeiat Raids/'oL

53. No. 193 /'ednesday, Octol05, 1988 / Notices previously approved on the RBS docket with respect to transient results of OLMCPR, MCPR, and MCPR<. ATWS. vessel overpressurization, peak clad temperature during a LOCA, annulus pressurization loads during a LOCA. reactor core thermal-hydraulic stability. and feedwater piping fatigue usage factor.

The acoustic and flow-induced loads on reactor internals created during a LOCA with FWHOS operation would increase slightly:

however, the results of the evaluation concluded that there is adequate conservatism In the evaluation and significant design margin remains available to account for these loads.

With respect to impact ofFWHOS on containment LOCA response, the peak drywell.to-wetwell differential pressure for the recirculation line break case increased by 1.02 psi. This differential pressure ls still considerably less and the design differential pressure of 25 psid presented in USAR Table 6.2-1. Also, the pool swell, condensatlon oscillation, and chugging loads evaluated at the worst power/liow condition during FWHOS operation increase slightly over the peak values presented in USAR Section B.

The analysis concluded this increase is insigniflcant and that adequate design margin exists to account for these loads.

The fatigue usage factor for the feedwater nozzle during FWHOS operation increased

'y 0.0214 over 40 years of continuous FWHOS operation assuming an 60% capacity factor. However, the fatigue usage factor would still be less than 0.6. which ls below the limitof 1.0.

The number of days of FWHOS operation must be limited to ensure that the feedwater sparger fatigue usage factor does not exceed 1.0. Administrative controls to ensure the number of days and magnitude of temperature reduction during FWHOS operation is tracked willensure that FWHOS operation cannot decrease the margin of safety as deflned In the bases to any Technical Specification.

Conservatism In the current TFSP Technical Speciflcation setpoint for the 40%

rated thermal power bypass of reactor scram on turbine stop valve position and turbine control valve fast closure and bypass of EOC RPT adequately accounts for FWHOS operation. Therefore, since the setpoint is unchanged. there is no reduction in the margin of safety for this setpoint.

The proposed change In Technical Speciiication Table 3.3.6-2. Item 1.b, High Power Setpoint. Is consistent with current design bases. Additionally, the proposed change is consistent with the conditions assumed In the RWE analysis and Is consistent with the upper range currently specified. The proposed change is also consistent with the Standard Technical Speciflcations and the Technical Specifications of other licensed BWR/6 plants.

It is thus concluded that FWHOS operation and the proposed change to high power setpoint Technical Specification do not

. reduce the margin ol safety.

In conclusion. the proposed operating change willnot Increase the possibility or the consequences of a previously evaluated event and willnot create a new or different kind of accident from any previously evaluated. Also. the results of this request are within all acceptable criteria with respect to system components and design requirements.

The ability to perform as described in the USAR is maintained and therefore. the proposed change does not involve a significant reduction in the margin of safety.

Therefore. GSU proposes that no significant hazards are involved.

The staff has reviewed the licensee's no significant hazards consideration determination. Based on the review and the above discussions.

the staff proposes to determine that the proposed changes do not involve a significant hazards consideration.

Local Public Document Room location: Government Documents Department, Louisiana State University, Baton Rouge, Louisiana 70803 Attarneyforlicensee: Troy B. Conner.

Jr., Esq., Conner and Wetterhahn, 1747 Pennsylvania Avenue, NW.,

Washington, DC 20008 NRC Project Director: Jose A, Calvo Louisiana Power and Light Company, Docket No, 50482, Waterford Steam Electric Station, Unit 3, St. Charles Parish, Louisiana Date ofamendment request/ August 26, 1988 Description ofamendmenl request:

The proposed amendment would change the Technical Specification by correcting the labels for percent level corresponding to Boric Acid Makeup Tank volume.

Basis forproposed no significant hazards cansideration determination:

The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92. A proposed amendment to an operating license for a facilityinvolves no significant hazards consideration ifoperation of the facility in accordance with a proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety. The proposed change is to correct the percent level which is read from the control room instruments and which corresponds to the volume in the Boric Acid Makeup Tank. The licensee determined that a quantity of water in the bottom of the tanks was not available for the pumps because. of the tank configuration. Additional water was added to the tank to make up the difference that waa not available and this water addition resulted in a change in the percents readouts in the control room; An analysis of the water required for analyzed accidents indicates there is more water than required and the Technical Specification chart for acceptable operation remains over conservative. The water addition and change to correct the corresponding percents does not involve a significant increase in the probabilities or consequences of any previously analyzed accident nor do they create the possibility of a new or different kind of accident. For the actual water available for reactivity control. the addition of water and change ofpercents does not involve a significant@eduction in a margin ofsafety. Based on the above.

the staff proposes to determine that the change does not involve a significant hazards consideration.

Local Public Document Room Location: University of New Orleans Library. Louisiana Collection, Lakefront.

New Orleans. Louisiana 70122 Attorneyforlicensee: Bruce W.

Churchill, Esq., Shaw, Pittman, Potts and Trowbridge, 2300 N StNW..

Washington, DC 20037 NRC Project Director: Jose A. Calvo Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point Nuclear Station, Unit No. 1, Oswego County. New York Date ofamendment raquest: March 28.

1988 Description ofamendment raquest:

The proposed changes to Table 6.2-1 would make the Table consistent with the requirements of10 CFR 50.54(m)(2)(i) for the minimum licensed operator staffing. and would provide additional clarification of the staffing required during hot shutdown versus that required during cold shutdown and refueling. The actual shift staffing would not change.

The proposed revision to Note 7 in Table 8.2-1 would make this Table consistent with the requirements of Technical Specification 8.2.2.e which requires the Assistant Station Shift Supervisor to assume the position of Shift Technical Advisor ifthe emergency plan ia activated during normal operations or hot shutdown.

Basis forprapased no signifiaant hazards cansideratian determination:

The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92. A proposed amendment to an operating license for a facilityinvolves no significant hazards considerations ifoperation of the facility in accordance with a proposed amendment would not: (1) involve a signiiicant increase in the probability or

yer>srs> ReOr / Vel 53,'Na 2!r>'/ Wedeesdsy, Oer S, 226> / Net/ees 38173 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 proposed changes willnot (1) involve a significant increase in the probability or consequences of an

,accident. (2) create the possibility of a new or different kind of an accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety for the following reasons; one change would merely indicate that at least two licensed operators must be on shift during hot shutdown and one during cold shutdown or refueling, as has been the practice in accordance with the regulations. This change is administrative since there is no change in actual shift staffing. The other change is.also administrative because lt would make the wording of Note (7) ofTable 6.2-1 consistent with existing Specification 82.2.e.

Based upon the above considerations, the staff prop'oses to determine that the proposed changes do not constitute a signiTicant hazards consideration.

LocalPublic Document Room location: Reference and Documents Department. Penfield Library. State University ofNew York, Oswego. New York 13128.

Attorne>rforlicensee: Troy B. Conner, Jr. ~ Esquire, Conner 8 Wet terhahn, Suite 1050. 1747 Pennsylvania Avenue, NW.,

Washington, DC 20006.

NRC Project Director: Robert A.

Capra. Director Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone Nuclear Power Station, Unit No. 3, New London County, Connecticut Date ofomendment request:

September 2. 1988 Description ofamendment request:

The proposed amendment would change Technical Specificatioa 3/4.23.2, "RCS Flow Rate and Nuclear Enthalpy Rise Hot Channel Factors Four Loops Operating," and TS 3/4.2.3.2. "RCS How Rate and Nuclear Enthalpy Rise Hot Channel FactorThree Loops Opera ting." The change would incorporate the followingrequirement in TS 4.2.3.1.6 and 4.2.34.6:

"Ifthe feedwater venturis are not inspected and cleaned at least once per 18 months. an additional 0.1% willbe addedto the total RCS flow measurement uncertainty."

Basis forpropose no significant hazards consideratton determination:

On January 20,1988, the NRC issued Amendment No. 12 to the Facility Operating License for Millstone Unit No.

3. Enclosure 2 to'he January 20, 1988 letter provided an evaluation ofthe licensee's methodology for determining reactor coolant system (RCS) flow. One component of the overalf RCS flow uncertainty is the uncertainty related to the condition of the feedwater flow sensing instrumentation. Since the feedwater flowventurl seasors are prone to fouling, overall RCS flow uncertainty may be increased by as much as.1% ifsuch fouling is not corrected. In the event that the feedwater flowventuri sensors cannot be inspected during refueling outages. it is conservative to assume that fouling has occurred and that the increase of

.1% for RCS flow uncertainty is applicable. Regarding the effect of venturi fouling on RCS flowuncertainty, to the NRC stafFs January

20. 1988 letter concludes:

TS sections 4.2.3.1.8. 4.2.3,2.8 and the bases for TS section 3/4.2.4 (page B 3/4 2-8) will need io be modified to state that Ihe penalty for undetected fo>>ling of the feedwater venturis of 0.1% willbe added to the flow measurement uncertainty values ifthe veniuris are not deaned. This is to be done before the precision heat balance is made to calibrate the RCS How rate indicators (approximately once per 18 months). The licensee has stated that the feedwater venturis have been cleaned for the Cyde "

operation.'I2>e licensee has stated (Ref. 20) that the above TFs willbe modified to reflect the requirement of 0.2% penalty ifthe van!uris are not cleaned and submitted for NRC approval. The staff requiie this.

modification prior to Cycle 3 operation.

At the present time, TS 4.2.3.1.6 requires that, in the event that the venturis are not inspected. the.l%

uncertainty factor for RCS flow is imposed. No "deaniag" requirement is contained in TS 4.2.3.1.8; however, the proposed change to TS 4.2.3.1.8 contains the deaning requirement. No similar requirement is presently in TS 4~2; however, the proposed change to TS 42.3.2.6 is identical to that proposed for TS 4.2.3.1.6.

On March 8. 1988. the NRC published guidance in the Federal Register (51 FR 7751) concerning examples of amendments that are not likely to involve a significant hazards consideration. One example of amendments not likelyto involve significant hazards considerations is example (ii)which involves, "Achange that constitutes an additional limitation, restriction, or control not presently.

included in the Technical Specifications, e.g., a more stringent surveillance.

requirement "The proposed changes to TS 4.2.3.1'and T84.2.3.2 are consistent with Example (ii) ln that they add an additional restriction, the imposition of a.1% flow uncertainty in the event that the venturis are not inspected and cleaned once per 18 months.

Accordingly. the staff has made a proposed determination that the application for.amendment involves no significant hazards consideration.

LocolPublic Docunient Room locotion: Waterford PuMic Library. 49 Rope Perry Road, Waterford, Connecticut 06385.

Attorneyforlicensee: Gerald Garfield, Esquire. Day. BeiTy 8> Howard. Ona Constitution Plaza. Hartford, Connecticut 08103-3499.

NRC Project Director: John F. Stolz

'orthern States Power Company, Docket No. 50-263. Moaticello Nudear Geaerating Plant, Wright County, Minnesota Date ofomendment request: February 16, 1987, as revised August 27. 1988.

Description ofomendment request:

The proposed license amendment involves various plant Technical Specificatlon (TS) changes to reflect changes to standards. guidelines. NRC administrative requirements, and to provide consisteacy with past data.

Changes initiallyproposed by letter dated February 16, 1987. to reflect changes in management titles and organization specified in Section 6 of the TSs were withdrawn b'y letter dated August 17, 1988. The proposed TS changes are as follows:

(1) Replace "... and 1 sample from a control location 8-20 miles distance and in the least prevalent wind direction".

with"... and 1 sample from a control location specified in the ODCM" in Item (1), Airborne Radioiodine and Particulates. ofTable 4.18.1 (Page 1 of 5) in the TSs.

(2) Delete page 251a from the TSs, incorrectly retained, and which should have been deleted by License Amendment No. 46 (July 1, 1986).

(3) Delete Section 8.8 from the TSs which was superseded by the publication of 20 CPR 50.49, "Environmental qualification ofelectric equipment important to safety for nuclear power plants."

(4) Standardize reports and correspondence to conform to 20 CFR 50.4 as follows:

(a) Replace "Director of the appropriate Regional Office of

, Inspection and Enforcement" with "US.

Nuclear Regulatory Commission, ATTN:

Document Control Desk, Washington, DC 20555."

(b) Delete "to the Office of Management Information and Program Control. U.S. Nuclear Regulatory Commission, Washington, DC 20>55."

39174 Federal Reglst Vol. 33, No. 193 / Wednesday,'ctobO, 1988./ Notices (c) Delete 'The reports listed below

'shall be submitted to the administrator of the appropriate Regional Office or designate."

(d) Delete "Written reports for the followingitems shall be submitted to the appropriate Regional Administrator."

(e) Delete "to the appropriate NRC Regional Administrator."

(5) Replace "Paragraph 4.4 of ANSI N18.7-1972" with "ANSIN18.7-1978 as modifled by the Operational Quality Assurance Plan" to reflect current ANSI standards referenced in the updated Quality Assurance Plan for Monticello operation.

Basis forproposed no significant

hazards consideration determination:

The Commission has provided standards for determining whether a significant hazards consideration exists (10 CFR 50.92(c)). A proposed amendment to an operating license for a facilityinvolves no significant hazards consideration ifoperation 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 licensee has evaluated the proposed changes against the above standards as required by 10 CFR 50.91(a) and has concluded that the proposed changes involve no signiTicant hazards consideration. The Commission has reviewed the licensee's evaluation and agrees with it for the following reasons:

Change (1) clariTies the location of the control for monitoring airborne radioiodine and particulates. The change in no way alters the control sample location as set forth in the approved Offsite Dose Calculation Manual (ODCM) or alters the intent of the TS requirements relative to the environmental monitoring program; no physical or procedural changes are involved; and it does not reduce the level of protection provided to the environment. The change only affects the way in which the location of an environmental monitoring program control sample is specified: i.e.. the change permits some flexibilityin obtaining a control sample. taking into consideration changes in wind patterns which vary from year to year. The literal interpretation of the existing requirement could unnecessarily require a change in sample location each year corresponding to variable wind patterns.

As such, this change is considered not to involve a significant increase in the probability or consequences of an accident previously evaluated, or to create the possibility of a new or different kind of accident from any accident previously evaluated, or to involve a significant reduction in a margin of safety.

Changes (2) and (3) are considered to be purely administrative in nature as documented by way of Example (i) published in the Federal Register (51 FR 7751); I.e., the changes achieve consistency throughout the TSs, correct an error. or change nomenclature.

Change (4) is considered to be applicable to Exemple (vii)published In the Federal Register (51 FR 7751) in that it is a change to conform a license to changes in the regulations (namely 10 CFR 50.4), where the license change results in very minor changes to facility operations clearly in keeping with the regulations.

Change (5) updates the TS auditing requirements reflecting a change in the ANSI N18.7 standard from 1972 to 1976.

ANSI N18.7-1978 is more stringent than the currently specified ANSI N18.7-1972 standard. and this change would incorporate the more stringent standard in the TSs. This change fits Example (ii).

published In the Federal Register (51 FR 7751) ~ since it is a change that constitutes an additional limitation, restriction or control not presently included in the TSs.'ased on the above, the Commission proposes to determine that the proposed amendment does not involve a significant hazards consideration.

Local Public Document Room location: Minneapolis Public Library.

Technology and Science Department, 300 Nicollet Mall, Minneapolis, Minnesota 55401.

Attorneyforlicensee: Gerald Charnoff. Esq., Shaw. Pittman, Potts and Trowbridge. 2300 N Street. NW.,

Washington. DC 20037.

NRCProject Director. Martin J.

Virgilio.

Northern States Power Company, Docket No. 50-263, Monticello Nuclear Generating Plant, Wright County, Minnesota Dote ofamendment request: August 14, 1987, as revised January 4, 1988.

February 10, 1988, and August 311988.

Description ofomendment request:

The changes proposed to the plant Technical Specifications (TSs) would:

(1) remove the figures in Section 8 depicting corporate and plant organizational charts and specify in lieu thereof general requirements that capture the essential aspects of the organizational structure that are defined by existing onsite and offsite organization charts. in accordance with the guidance provided In NRC Generic Letter No. 88-08 (March 22, 1988); and (2) delete the requirement for plant management and support staff not assigned to a rotating operations shift to hold a current Senior Reactor Operator (SRO) license.

Basis forproposed no significant hazards consideration determination:

The Commission has provided standards for determining whether a significant hazards consideration exists (10 CFR 50.92(c)). A proposed amendment to an operating license for a facilityinvolves no significant hazards consideration ifoperation 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 licensee has evaluated the proposed changes against the above standards as required by 10 CFR 50.91(a). The Commission's staff has reviewed the licensee's evaluation and agrees with it. The licensee concluded that:

1. The changes proposed to remove corporate and plant organization charts from the TSs do not involve a signiTicant increase in the probability or-consequences ofan accident previously evaluated or create the possibility of a new or different kind of accident from any accident previously evaluated. or involve a significant reduction in a margin of safety. As stated in NRC Generic Letter No. 88-06. the requirements necessary for safe operation of the plant have been retained in the TSs: the changes do not eliminate or alter the functions previously reviewed; and the changes do not affect plant operation and design or create a new accident mode. The changes proposed were modeled after to NRC Generic Letter No.

88-08 in conformance with Commission requirements.

2. The changes proposed to eliminate certain requirements for plant managementnnd support staff to hold current SRO licenses do not involve a significant increase in the probability or consequences of an accident previously evaluated because there are no changes being made to the license requirements for individuals controlling the reactor and other plant systems; there willbe no impact on the quality of plant operatio'ns. and therefore. the changes willnot result in a degradation in plant

Federal Reg

/ vol. oo, No. 19$ J'Wedoeedey, Oct 5, 1988:/ Notices operations which would increase the probability ofan accident. No changes are proposed in the license requirements for personnel actually operating the

'eactor and other plant systems or shift management which would create the possibility of a new or different kind of accident from any accident previously evaluated, or involve a significant reduction in a margin of safety.

Accordingly, the Commission proposes to determine that the proposed amendment does not involve a significant hazards consideration.

Local Public Document Room lacatian: Minneapolis Public Library.

Technology and Science Department, 300 Nicollet Mall,Minneapolis, Minnesota 55401.

Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Pot ts and Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director; Martin J.

Virgilio.

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

September 2, 1988 Description ofamendment request:

This proposed amendment would revise the Technical Specifications (TS) to support Cycle 12 operation. The proposed amendment would modify the TS as follows:

(1) TS's 1.1 and 2.10.4(3) would be changed to reduce the calculated value for the limitof the Total Unrodded Planar Radial Peaking Factor (F~T) to 1.80 which willprovide additional operating margin. The correction of a Cycle 11 setpoint evaluation. in the use of the more limitingLoss of Coolant Accident Required Overpower Margin (ROPM) versus the transient analysis ROPM, has reduced the "tent" for the core power limitversus the AxialShape Index (ASI) for the LimitingCondition for Operation (LCO) for Excore Monitoring of the Linear Heat Rat (LHR), Figure 2-6. By reducing F T, additional operating margin is gained in this LHR-LCO operating tent. Figure 2-9 would be changed such that the Total Integrated Radial Peaking Factor and the Total Planar Radial Peaking Factor limits are consistent with the change in Figure 2-6.

(2) TS 2.10.4(1) would be changed to provide clarification as to how the linear heat rate should be monitored and what parameters apply to bound the limits. In particular. the point at which the limitingcondition for continued operation without reducing power, should the plant computer incore detector alarms become inoperable, is clarified as seven days from the date of the last valid core power distribution.

Also, the requirements for.maintaining the AxialShape Index, Yn within the limits of Figure 2-6 when linear heat rate is continuously being monitored by excore detectors, are clarified.

(3) Figure 1-3 and TS 2.10.4(5) and Footnote " on page 2-57c would be changed to reduce the limitfor Cold Leg Temperature to 543'., indicated, and 545'., actual, from 545'. and 547'.,

respectively. This is being done to reflect actual operating conditions and to gain margin. It.has also resulted in changes to the Alpha, Beta, and Gamma terms of the Thermal Margin/Low Pressure equation for Figure 1-3.

(4) Administrative changes that would correct a typographical error in TS 1,3(1) and change references to the final safety analysis report (FSAR) to the updated safety analysis report (USAR).

Basis forproposed na significant hazards cansideratian determination:

The Commission has provided standards for determining whether a significant hazards consideration exists as stated in10 CFR 50.92(c). A proposed amendment to an operating license for a facilityinvolves no significant hazards consideration ifoperation 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; (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 licensee addressed the above three standards in the amendment application.

(1) Reduction of the Total Unrodded Planar Radial Peaking Factor, F T, from 1.85 to 1.80. With regard to the three standards, the licensee states that operation of the'facility in accordance with this amendment would not:

(a) Involve a signiAcant Increase in the probability or consequences of an accident previously evaluated. This change merely allows utilization of the additional margin available with the reduction ofmaximum F T value with no changes in administrative specifications. On the basis of technical safety evaluation. operating with gain in margin for Cycle 12 LHR-LCO would be no more limitingthan operating with the Cycle 11 LHR-LCO. Therefore, this change does not increase the probability or consequences of a previously evaluated accident; ib) Create the possibility of a new or different kind of accident from any accident previously evaluated. It has been determined that a new or different type of accident is not created because no new or different modes of operation are proposed for the plant. The continued use of the same Technical Specification administration controls prevents the possibility of a new or diffecent kind of accident.

(c) Involve a significant reduction in a margin of safety. Administrative specigcations involving the LHR-LCO ensure that operating with the extra margin gained'rom the reduction ofF T conforms to current plant conditions and, therefore, preserves the margin of safety. Reducing the LHR-LCO tent does not affect the available

'argin and, therefore, willnot reduce the margin of safety.

(2) Decrease the cold leg temperature from 545'. to 543'. With regard to the three standards, the licensee states that operation of the facilityin accordance with the amendmettt would not:

(a) Involve a signilicant increase in the probability or consequences of an accident previously evaluated. This change allows the reduction ofT, to 543'. The temperature change Is bounded by the previous technical safety analysis which addressed the 545' inlet temperature. Therefore. this change does not increase the probability or consequences of a previously evaluated accident.

lb) Create the possibility of a new or different kind of accident from any accident previously evaluated. It has been determined that a new or different kind of accident Is not created because no new or different modes of operation are proposed for the plant. The continued use of the same Technical Specigication administrative controls prevents the possibility of a new or different kind of accident.

(c) Involve a signiAicant reduction in a margin ofsafety. Administrative speci Acations involvingT, ensure that operating at a T, of543' conforms to current plant conditions and. therefore, preserves the margin of safety. The temperature change Is bounded by previous technical safety analysis which addressed the 545' inlet temperature and. therefore, willnot reduce the margin of safety.

(3) Changes to the instructions for the entering of the LimitingCondition for ~

Operation (LCO) for Excore Monitoring of Linear Heat Rate (LHR), With regard to the three s'tandards.

the licensee states that operation of the facilityin accordance with the amendment would not:

(a) Involve a significant increase in the probability or consequences of an accident previously evaluated. This change clariAes the point at which the LHR-LCO (Figure 2-8) must be entered and provides better guidance forplant operation. The basis for the technical safety evaluation would be no more limitingthan operating with the Cycle 11 basis. Therefore, this change does not Increase the probability"or consequences of a previously evaluated accident.

ib) Create the possibility of a new or different kind of accident from any accident previously evaluated. It has been determined that a new or different type of accident Is not created because no new or different modes of operation are proposed for the plant. The continued used of the Technical Specification administrative contiols prevents the

issdelei Re'iyiseveL'59, Ne; eely 'f'llVeiMeece'y, achkeO \\'9!ss I Neiices possibility of~ new or d'Ifecent kind of accident.

(c) involve ci eigniTicsnt redaction ia a margin of safety. Administriitive eyedfications involving the LHR-LCDeneace that the operators eater ihe LCO with sufficient time to reduce power. ifiiecewiary, prior to utilizing the excore Inetcuauyiite co monitor core power. The changes have been implemented through strict sdnintstrative procedures and. therefore, willaoi reduce the margIn ofsafety.

The staff has reviewed the licensee's no significant hazards consideration determination and agrees with the analysis.

The Commission has provided guidance concerning the application of the standards for determining whether a significant hazards considerafion exists by providing examples (51 FR 7751) of amendments that are considered not likelyto involve significant hazards consideration. The proposed administrative changes (item 4) in this amendment are similar to the example of a purely administrative change to the Technical Specifications. Accordingly, the staff proposes to determine that the proposed changes to the Technical Specification do not involve a significant hazards consideration.

Local Public Document Room locatioru W. Dale Clark Library. 215 South 15th Street, Omaha, Nebraska 68102 Attorneyforlicensee LeBoeuf. Lamb, Leiby. and MacRae.1333 New Hampshire Avenue, NW., Washhigton, DC 20038 NRC Pxoject Dictator: Jose A. Calvo Phiiade)phia Electric Company, Docket No. 50-352, Limerick Generating Station, Unit 1, Montgomery County, Pennsylvania Date ofamendment request: July 19, 1988 Description ofamendment request:

The proposed amendment responds to guidance provided in the stafFs Generic Letter 87-09 dated June 4, 1987.

Specifically, the proposed amendment would modify the general limiting conditions for operation (LCO) to allow entry into an operational condition under certain circumstances when compliance with the LCO's related Action Statements would allow continued operation for an un)imlted period of time.'Ice general surveillance requirements would abo be modified to clarify the time at which Action Statement time limits begin relative to failure to perform a surveillance requirement and to allow for a de)ay of the Action Statement requirements for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to complete the'urveillance ifthe a))oweble ticne is less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. It would a)so c(arify that restrictions on entry ntto~thna)

Conditions baeed on faihcie to comply with surveillance requireinents shall not prevent passage into or through Operational Conditions as mpired by Action Statements. Ae re)atacd bases have also been changed to refiect the proposed changes to thiTechnical Specifications (TS).

In addition, the amendment deletes numerous TS statements which presently take exception to the provisions ofTechnical Specification 3.0.4.

Basis forproposed na significant hazards considera tion detenninati ant On June 4, 1987, the staff issued Generic Letter 87-09, "Sections LO and 4.0 of the Standard Technical Specifications (STS) on the applicability oflimiting conditions for operation and surveillance requirements." Thatletter contained guidance for improvement of Sections 3.0 and 4.0 of the STS consistent with the recommendatlons of NUREG-1024, "Technical Specifications Enhancing the Safety Impact." and the Commission's Po)icy Statement on Technical SpeciTication improvements. The licensee's submittal conforms to the staffs guidance.

The licensee has provided an ana)ysis as to whether the proposed amendment involves a significant hazards consideration. The licensee's analysis is summarized as follows:

The standards used to arrive at a determhiattoci that a request for amendment requires no significant hazards coneMeration are Indaded In the Commission'i Regulatians. 10 CFR $0.92. which state that the operation of ibe fadliiyin accordance with the proposed amendment wou)d not (1) involve a significant increase Ia the probability or coaeeqaenoei of an acdcient previously evaluated, (2) create the possibility of a new or dtfferent kind of acddent from any ecddent previoaeiy evaluated, or(8) Involve a eigntficant redaction in a caargtn ofsafety.

The proposed changes do not Involve a sfgaificaat Increase tn the probability or consequences of an aoddeat previoualy evaluatect The changes being proposed are admiiitetrative in nature and are being made to correct hconeieteadee In ibe present wording of tbe 8eaeral Sections Nland 4.0 of the Technical Spedfications. As suck tbe proposed changes do not affect any evaluated acddent.

The proposed changes do not create the possibility of a new or diiferent kind of ecddent. As stated above, the proposed changes are adiaiaistrative changes whkh do not create the poasibl)tty ofany new accident.

The proposed changes do not involve a significant reduction In the margin of safety. The changes to Sectioa 30.4 allow starbips under conditions

~

whereby conformance to the Action Requirements establishes an acceptable leva( ofsafety foranlimited continued operation of the feei)ity, while de)eying a return to poweraperation when the facilityis required to be shut down as a conseqhence af an Action Requirement.

The change to Section 4.0.3 afiows appropriate time for performing a missed surveillance before shutdown requirements app)y to permit the performance of the missed survel))anes based on consideration of plant conditions, adequate planning, availability of personnel. and the time to perform the surveillance. The NRC staff stated in the Generic Letter that it is overly conservative to assume that systems or components are inoperable when a surveillance has not been

'erformed.

Therefore, allowing sufficient time to perform the surveillance does not eignificantiy reduce the margins of safety.

The final change to Section 4.0.4 Is a clarification to permit passage through or to operational modes as required to comply with Action Requirements even tboagh a surveillance requirement bas not been performed. To nat permit this would increase the potentia) forplant apsets, and woald challenge safety eystemL The revision woald also permit mode changes when a surveillance requirement has not been mei, and can only be completed after entering into a mode or specific condition. 'I1ils condition does not significantly reduce the margin af safety, but In fact potentially Increases the raargin of safety, by permit ting entry into lower modes ofoperation more quideiy. Thus.

there is not a significant reduction tn the margbi of safety.

The s taIfhas reviewed the licensee's submitta) and significant hazards analysis and.has determined that the proposed Technical Specifications conform to the staff guidance contained in Generic Letter 87-09. Farther, the staff concurs with the )Icensee's determination as to whether. the proposed amendment involves a significant hazards consideration.

Therefore, the staff proposes to determine that the proposed amendment involves no significant hazards consideration.

Local Public Document Room lacatiam Pottstown Public Library, 500 High Street, Pottstown, Pennsylvania 19464.

Attorneyforlicensee: Conner and Wetterhahn. 1/47 Pennsylvania Avenue, NW., Washington, DC 20006 NRC Prat'ect Director." Walter R; Butler Pawar Authority ofthe State of New York, Docket Noe 50-333, Jacques A.

FitzPatrick Nuc)aar Power Plant.

Oswego, New York Date ofamendment request: August 24, 1988

yedesal Reg

/ Vol: 33, No. 133 / Wednesday, Oct 5, 1988 / Notices 39

Description afamendment request:

The proposed amendment would revise the Technical Specifications (TS) to reflect modifications made to the Standby Liquid Control System (SLCS) during the Reload 8/Cycle 9 refueling outage. In accordance with the requirements of 10 CFR 50.82, changes are being made to the SLCS to ensure a minimum flow capacity and boron content equivalent to 68 gallons per minute of 13 weight percent sodium pentaborate solution. In addition to meeting 10 CFR S0.82 requirements, the final in-vessel boron concentration followinginjection of standby liquid control solution is being increased to permit an increase in fuel'reload enrichment and eneqp content in future core design.

Basis forproposed na significant hazards consideration determinatian: In accordance with the Commission's Regulation in 10 CFR S0.92, the Commission has made a determination that the proposed amendment involves no significant hazards considerations, To make this determination the staff must establish that operation 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 proposed modification to the SLCS involve an increase in B-10 enrichment in the solution in the SLC Tank and an increase in required pumping capacity. Although the modifications involve decreasing the concentration of sodium pentaborate in the SLC Tank. the increased enrichment of B-10 and increased solution pumping rate result in an overall increase in the injection rate ofB-10 isotope into the reactor vessel. As a result of the increased amount ofB-10 isotope In the SLC tank, the final in-vessel boron concentration followingInjection of SLC solution is being increased from 600 ppm of natural boron to 860 ppm of equivalent natural boron. The increased boron concentration in the reactor vessel willallow future core reloads to utilize higher energy content fuel without decreasing the present shutdown margin. Furthermore, operation of the SLCS with the proposed changes willmerely provide a backup to other safety-related systems in accordance with 10 CFR 50.62 requirements and willnot affect any previously analyzed accidents.

Therefore. operation of FitzPatrick in accordance with the proposed amendment does not involve a significant increase In the probabfllty or consequences ofan accident previously evaluated.

The SLCS serves as a backup to already existing safety-related systems.

The proposed modifications willensure that the SLCS is maintained such that it is capable of fulfillingthe operability requirements of 10 CFR 50.62. As stated above, the proposed changes increase the shutdown margin in the unlikely event that SLCS should be needed to shut down the reactor. No new or

'ifferent kinds of accidents result from improving the effectiveness of the SLCS.

Therefore, operation ofFitzPatrick in accordance with the proposed

" amendment willnot create the possibility of a new or different kind of accident from any previously evaluated.

The proposed modifications increase the negative reactivity inserted by the SLCS, and, therefore, enhance the safety margin for the plant. The proposed changes are intended to meet with the requirements of 10 CFR 50.62, and provide additional assurance that the SLCS is capable of safely shutting down the plant in the unlikely event that its use is required. Therefore, operation of FitzPatrick in accordance with the proposed amendment does not involve a significant reduction in a margin of safety.

Since the application for amendment involves proposed changes that are encompassed by the criteria for which no significant hazards consideration exists. the staff has made a proposed determination that the application involves no significant hazards consideration.

Local Public Document Boom location: State University ofNew York, Penfield Library, Reference and Documents Department, Oswego, New York 13126.

Attorneyforlicensee: Mr. Charles M.

Pratt. 10 Columbus Circle, New York, New York 10019.

NBCProject Director: Robert A.

Capra, Director Tennessee Valley Authority, Docket No.

50-260, Browns Ferry Nuclear Plant, Unit 2, Lbnestone County, Alabama Date ofamendment requests: August 3, 1968 (TS 250)

Description ofamendment requests:

The Tennessee Valley Authority (TVA) has proposed changes to the Browne Ferry Nuclear Plant, Unit 2 Technical Specifications (TS). The proposed changes are to incorporate surveillance fequlrenlellts isild trip level settings for new temperature switches being installed near a pipe trench containing Reactor Water Cleanup (RWCU) System piping. The added instrumentation will indicate leaks or pipe breaks and automatically isolate the RWCU system piping.

Basis forproposed no significant hazards cansideratian determination:

The Commission has provided Standards for determining whether a significant hazards determination exists as stated in 10 CFR 50.92(c). 10 CFR 50.91 requires that at the time a licensee requests an amendment. it must provide to the Commission its analyses, using the standards in Section 50.92, on the issue of no signific(tnt hazards consideration. Therefore, in accordance with 10 CFR 50.91 and 10 CFR 50.92, the licensee has performed and provided the followinganalysis:

NRC has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92(c). A proposed amendment to an operating license involves no significant hazards considerations ifoperation of the facilityIn 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 an accident previously evaluated, or (3) involve a significant reduction ln a margin of safety.

1. The proposed amendment does not Involve a signilicant!ncrease in the probability or consequences of an accident previously evaluated because the amendment only adds operability surveillance requirements and trip level settings for new temperature detectors. The Final Safety Analysis Report specIAes that the trip level setting be high enough to avoid spurious operation but low enough to prevent excessive loss of reactor coolant.

Establishing the trip level setting range of 130' to 150' satifies that requirement.

Establishing the same operability requirements on the new temperature switches as are on the presently Installed Instrumentation prevents a signiltcant increase in the probability or consequences of an accident previously evaluated. The system isolates for several accident conditions and since it serves no safety function. increasing the number of devices which could cause system isolation willnot affect safe operation of the plant.

2. The proposed amendment does not create the possibility of a new or different kind of accident from an accident previously

, evaluated. The new temperature switches are performing a similar function as other instrumentation presently installed, and setting their operability and surveillance requirements the same as presently Installed temperature switches prevents the creation of a new or different kind of accident. The Increased monitoring and automatic isolation for the RWCU System willhelp prevent damage by high teinperature to equipment required for safe shutdown. The addition of operability and setpoint requirements for the

new temperature switches ensures Ibat the system's pcicaacy contataaient isolation safely fixictfonwillbe performed adequately.

The change does not affect safety functions of any equipment In ways not previously evaluated.

3. The proposed amendment does not involve a signifliceat ceductfcui fa a mscgfn of safety because tbe temperature switches being added are being specified to meet the same requirements as other RWCU Syalcm temperature switches which perform the same function end are already tnduded in the technical specfflcatfena. This ensures that the new temperature switches willnot degrade exfatfcig features included in the technical apecfficatfons. Aiao. tfie new temperature switches are being added to eaauce that safety-related equfpmeat that Is addressed fn the technical spccfQcatfoaa and that is requfced lo mitigate s RWCV System pipe break wfiinot be degraded by the environmental conditions which could ceauit from a RWCU pipe break in the pipe trench.

The staff has reviewed the licensee's no significant hazards consideration determination and agrees with the licensee's analysis. Therefore. the staff proposes to detecmine that the application for amcndcnents involves no significant hazards considerations..

Local Public Document Room lacalian: Athens Public Library, South Street. Athens, Alabama 35811.

Attorney forlicansea General Counsel, Tennessee Valley Authority, 400 West Summit HillDrive. E11 833, Knoxville,Tennessee 37902.

NRCAssistant Direatatr Suzanne Black Tennessee Valley Authority, Docket No, 50-327, Sequoyab Nuclear PlaaL Unit1, Hamiltoa County. Tennessee Dale ofamendment request:

September 21, 1988 (TS 88-28)

Description ofamendment raqccestc The Tennessee Valley Authority (TVA) proposes to modify the Sequoyah Nuclear Plant Unit 1 (SQN) Technical Specifications (TS). The change is to revise the limitingcondition for operation (LCO) 3~ and surveiuanca requirement (SR) 4.2.2.2 to reflect a reduction in the heat fluxhot channel factor (Fo(z)) limitfrom 2.237 to 2.15.

Basis forproposed no significant hazards consideration determination:

TVAprovided the followinginformation in its submittal on the requested change to the beat flux hot channel factor.

By letter dated August 15, 1M8, TVA submitted proposed license amendment 88-20. This proposed change revised the upper head injection (UHI) isolation setpoint and tolerances of SR 4.5.1.2.c.1. of the August 15 letter describes that. as part of the setpoint change. the delivered UHI water volume band was being expended from the range of 1.13IL5 to 900 cubic feet to the range of 1,MLS Ia 859cubicfeet. The change in the delivered UHIwater volume band was supported by Westinghouse Electric Corporation (W) evaluations. which indicated that the potential decrease ia delivered water volume to the core would result in increased peak clad temperatures (PCTs): but in all cases. PCT remained below the 2,220 degree Fahrenheit (F}

limitof 10 CFR $0.48.

In telephone conversations on September 1 and 2, 1988, NRC informed TVAthat the increased PCTs described in the August 1$. 1988 submittal could not be whollyjustified by the sensitivity studies provided. NRC stated that restart ofUnit 1 could be supported by the sensitivity studies (Provided a temporary exemption to certain administrative requirements of 10 CFR 50.46(a)(1) was obtained) and that operational restriction be imposed to provide at least 100 degrees F of margin between the calculated PCT and the 10 CFR SIL48 limit.

TVA's request for a temporary exemption to certain administrative requirements af 10 CFR 50.46(a)(1) (has been] transmitted by separate correspondence.

[dated September 19, 1988].

Evaluations by (W) have determined that at least 100 degrees F PCI'argin can be obtained by administratively limitingsteam generator tube plugging to 5 percent and by reducing Fa(z) from

'.237 to 2.15. This proposed technical specification change is being submitted to reflect the reduction ln the Fo(z) limIt.

The Commission has provided Standards for determining whether a significant hazards determination exists as stated in 10 CFR 50.92(c). 10 CFR 50 qI requhes that at the time a licensee requests an amendment, it must provide to the Commission its analyses. using the standards in Section 50.92, on the issue of no signiflcant hazards consideration. Therefore. in accordance with 10 CFR 50.91 and 10 CFR 50q2. the licensee provided tha followinganalysis:

TVAhas evaluated the proposed technical specification change and has determined that itdoes aot cepceaent a sfgniQcacn hazscda conaidecatioa based on ccftecfa eatabllabed ha 10 CFR 50.9Z(c). Operation ofSQN in accordance with the proposed amendment wfifaoa (1) Involve a sfgnificant Inxeasc In the prebsbfffty or cociaeqoences of an accident previously evaluated. Fo(zl Ia deflned aa the mexfmum local hest Qux on the siicfsce of a fuel cod divided by the core average hect flux. Fo(z) Is used to lfmft the magnitude of hot spots end fa used as a bounding Input for accident analysis. Pa(x) is noc postulated as being the fnftfetfag event for any accident scenario.%mcefoce, lbe proposed chacige does mt affect tha pcabebifity, ofany accident pceNfcmafy cvalustecL %be proposed reduction fa Fa(z) from X2Kto 2.15 Is conservative ia naacce, tci tfist It results fa reduced PCTa ducfng a postulated accident.

The Fo(x) reduction serves ca an operational restcfctfon to ensure that PCTa remain below the 10 CFR 50.48 ffmkof2,200 degrees P.'ecause of the cedectfen fa calcafeted PCT; the pcopoeed change wgl aot fncceeac lbe consequences ofa previoaafy evaluated accident.

(2) Ccecle thc posaibiflty ofa new m differeat kind of accfdecit fcom any previously anafyzecL As slated above, Fo(x) fa not assumed to be the fnitiaucig event for any accident scenario. The proposed change lo F<(z) pcevfdes additfonal PCT mscgin to ensure that the X200degceea P Hmft fa not exceeded. The pceaeoce ofadditional margin willnot cceale the Posafbftfty ofa new or diffecent kind ofacxddeat.

(3) Involve a significant reducuoa ia a margin of safely. The proposed cedcictfon ia Fo(x) fa conservative ia aaticce as ft lowers the calculated PCI'or the Ifmtftlag LOCA analysis case. Aa cakxdsted by (WJ, the proposed redaction fa Pafz) fcoaa 2237 lo Z.15 lowers Oie calculated PCT by a7 degrees F for the fimftbigimperfect mfxfag case acid by aa degcees F forthe licaftiagpecfect mixing case.

These ceductfoaa. comfined with PCT margin obtained by sdmiaistcatively Itmitfag steam generator tube plugging to 5 peccent. result in calculated PCTs of2AMdegrees F for tbc limitingtmpecfecrmixing case aad 2,0N'egrees P far tbe ffmitfagpeceent mhfng case. Beccaae the cakafated PCT cemaina below the 2.220 degrees P limitof 10 CPR 50.40. thece fs ao cedsctkui fa tbe margin of safety lo cladding fadhcce. aad additional margin fs being added.

The staff has reviewed the hcensee's no significant hazards consideration determination and agrees with the licensee's analysis. Therefoce, the staff proposes to determine that tha application for amendments involves no significant hazards considerations.

LocalPublic Documint Boom location: Chattanoogakiamiltoa County Library, 1001 Broad Street. Chattanooga.

Tennessee 37402.

Attorney forlicensea Geaeral Counsel, Tecmesaa'e Vauey Authority, 400 West Summit HQI Drive, E11 833.

Knoxville.Tennessee 37902.

NBCAsscstant Dc'rector: Suzanne Black Tennessee VaIIey Authority, Docket Nos. 50-327 aad 50-328, Sequoyab Nuclear P)am'nits 1 and 2, HamiIton County, Tennessee Date ofamendacecct ceqcaestcc juae 2t.

1988 (TS 88-23)

Dasarr'ptiocc afamendncecct reqccesta.

'Ibe proposed acciaadaamt vvoald change the expiratfoa dates for the Operating License DPR-77 (Unit1) fcacn h!ay 27, 2010 to September 17, 2020 and for the Operating License DPR-79 (Unit 2) fcacn May27. 2010 to Sap tecaber 15, 2021.

Fedesai Rays / VoL1 SS. No. 199 /'ednesday. Oelr 9, 1999 / Noiinaa The current operating lioense expiration dates are 40 years from the date of issuance ofthe construction permits (May 27. 1970, for both units).

Since the Unit.1 full-power operating license was issued 10 years and 4 months after construction permit issuance (11 years and 4 months forUnit 2), the effective period of the Unit 1 license is approximately 29 years and 8 months (28 years and 8 months for Unit 2). The licensee's application requests a 40-year operating license term from the operating license for both the units because the units were designed and constructed on the basis of40-years of plant operation.

Basis forproposed no s/'gnifi'cant hazards considerntion determinntion:

The Commission has provided standards for determining whether a significant hazards determination exists as stated in 10 CFR 5(L92(c). 10 CFR 50.91 requires that at the time a licensee requests an amendmenL it must provide to the Commission its analysis. using the standards in Section 50.92, on the issue ofno significant hazards consideration.

Therefore. in accordance with 10 CFR 50.91 and 10 CFR 5L92, the licensee has performed and provided the following analysis and states that the operation of Sequoyah (SQN) in accordance with the proposed amendment willnot:

(1) Involve a sigiiificant increase ln the probability or consequences of an accident prevloualy evaluated. SQN Unit 1 and Unit 2 were dcsigsied end constructed on the basis of40-years ofpie/it operation. SQN's reactor vessel was fabricated and designed for a 40-year life. A comprehensive vessel materials surveillance program Is maintained in accordance with 10 CFR 50. Appendix ILAn analysis was performed to demonstrate compliance with tbe NRC pressurized thermal shock (IRIS) screening criteria in uccordaace with 10 CPR 5L01(b)(2). The assessment of the projected PIS reference temperature demonstrated that the SQN Unite 1 and 2 pressure vessels would meet the toughness requirements of 10 CPR 50M for 32 effective full-power years ofoperation which Is equivalent io a 40-year design life with an 80-percent capacity factor. Aging analyses have been performed for all safety-related electrical equipment in acconhmce within the scope of 10 CFR 5IL40 (harsh environment). The qualifuyd lifeof ihe equipment or compooeat Is tucoiporated within SQN's maintenance and replacement practices to ensiue that this safety-rc)ated electricc! equipment remains qualified and available io perform Iis safety function regardless of the overall age of the phmt.

Programs are iu place to detect abnormal deterioration aud aging ofcritical pleat components. These programs include A. ASME Boiler and Presiue Vessel Code.

Section Xi, and 10 CFR 50 Section 50~g).

1. In-Service Inspection [ISI) Prcygram-This piogram ensures that phiiit prcssure retaining vessels. piping. and support systems are inspected Iu aoconbnce with the ASME Section XIcode.

2. In-Service Test (IST) piogram-bahts program easures that suitety-related tiampa and valves are tested in accordance with the ASME Section XIcode.

B. Technical Specifications.

In addition to the SI and ST programs. the followingSQN technical specification also provide a means of monitoring the cumulative effects of power operation during the lifetime of the plant.

1. Specification 3.4S-Steam Generators-An augmented steam generator inmrvice inspection program demonstrates operability of SQN's steam generators over the life of the plant.

2. Specification 3.4.9.1-Reactor Coolant System Pressure/Temperature Limits-'IIie pressure and temperature of the reactor coolant system are limited to protect agaiast non.ductile failure of the reactor coolant system. These limits are updated periodically over ihe life of the plant io ensure that the fracture toughness requirements for the ferritic material within the reactor coolant pressure boundary are maintainesL

3. Specification 3.4.10-Reactor Coolant System Structural IntegrityThe ISI and IST programs. In conjunction with the additional inspections required for the Reactor Coolant Pump fiywhoet aiid reactor vessel nozzets.

ensures the s ructuial integrity aad operational readiness of these compoaents willbe maintained throughout the life af the p!ant.

4. Specificafion Sy.1-Component Cyc)ic or Transient LimitMonitoring. recording. aud evaluation of certain cyclic and transient limits provides a high level of confidence that certain components within the reactor coolant and secondary systems will not'xperienc fatigue feikue over their 40-year design life.

(2) Create the posaibtlity of a new or different kind of accident from any previously analyzed. The proposed amendment Is administrative in nature and does noi affect the safety analysis. plant equipment. or the physical facility. Because the accident analysis ofSQN's PSAR remains bounding. no new or different kiiidof accident scenarios ere created by this change.

(3) Involve a significant reductioa in a margin of safety. The proposed amendmeiit involves only a chaiige to the expiration dates of the operating licenses. Because SQN is based on a 40-year service life. this change willnoi affect the safety margtiis.

The staff has rev(esved the licensee's no significant hazards consideration determination and agrees with the licensee's analysis. The staff believes that existing programs in place to detect any abnormal deterioration and aging of critical plant components also prevent any signiTicant increase in the probability or consequences ofnn accident previously evaluated, or create the possibility of any new accidents. or any significant reduction in the margin of safety. Therefore, based on this review. the etaff has made a proposed determination that the application for amendments involves no siyuTicant hazards consideration.

LocalPublic Document Room location:,Chattanooga-Hami(ton County Libraiy, 1001 Broad Street. Chattanooga, Tennessee 37402.

Attorneyfor1icensee: General Counsel. Tennessee Valley Authority.

400 West Summit HillDiive, E11 B33, Knoxvi))e. Tennessee 37902.

NRC Assistant Director: Suzanne Black Yankee Atomic Electric Cnrupany Docket No. $8-029 Yankee Nuc)ear Power Station, Franklin County, Massa chasetts Date ofomendment request: Augusl 11, 1988 Description ofamendment request:

The proposed amendment wou)d delete references to specific values of boron concentration and to the requirement for an inverse count rate multiplication measurement under stated conditions.

Basis forproposed no significant hozards considerodan determination:

The Commission has provided standards for determining whether a significant hazards determination exists as stated in 10 CFR SIL92(c). Aproposed amendment to an operating license invohws no significant hazards considerations ifoperation of the foci)fty in accordance with the proposed amendment would noL (1) involve a significant increase in the probability or consequences of an accident previously evaluateth or (2) create the possibility of a new or different kind of accident from any accident previously eva)uatech or (3) involve a significant reduction m a margin ofsafety.

The licensee's analyses contained in the August 11, 1988 letter states the followin(p This change is requested in order io replace reference to a specific vahie with a more generalized fonu wfich willmeet the LCO requirement. and to delete a surveilhnce requirement tu Mode 8 which is unnecessary.

As such. this proposed change woidd rent:

l. Invoh e a significaiit increase ia the probability or coasequences ofan accident previously evaluated. This change willnot Increase signiiicantiy the piobability ar consequences ofan accideoL as the shutdown nuugin of ibe cere willcontinue io be adequately monitored and uurscteiit control to preclude Inadvertent criticality already exists.,

2. Create tbe possiMity of new or different kind of accident from any previously analyzed. This modificsttoa only provides an administrative wording change and deletes an unnecessary surveillance rcquiremenL Therefore, it does not create the passibility of a new or dlilerent ktud of an accident because tt does not modify plant operation.

~ 3S180 Federal Radiatelot.

SS. No. 199 / Wednesday. Octobe1999'/

Notices

3. Iovolve a significant reduction In a margin of safety. This modification only provides an administrative wording change and deletes an unnecessary surveillance requirement which does not affect the safety margins which currently exist during Mode 6 operation. Thus, this change does not Involve e reduction in a margin of safety.

Based on the considerations contained herein, Il Is concluded that there is reasonable assurance that operetion of the Yankee plant. consistent with the proposed Technical Specifications. willnot endanger the health end safety of the public. This proposed change hes been reviewed by the Nuclear Safety Audit end Review Committee.'he staff has reviewed the licensee's analysis and agrees with it. Therefore, we conclude that the amendm'ent satisfies the three criteria listed in 10 CFR 50.92. Based on that conclusion the staff proposes to make a no significant hazards consideration determination.

LocolPublic Document Room locotion: Greenfield Community College, 1 College Drive, Greenfield, Massacusetts 01301.

Attorneyforlicensee: Thomas Dignan, Esquire, Ropes and Gray. 225 Franklin Street. Boston, Massachusetts 02110.

NRC Prof ect Director: Richard H.

Wessman PREVIOUSLY PUBLISHED NOTICES OF CONSIDERATION OF ISSUANCE OF AMENDMENTS.TO OPERATING LICENSES AND PROPOSED NO SIGNIFICANTHAZARDS CONSIDERATION DETERMINATION

, ANDOPPORTUNITY FOR 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.

WolfCreek Nuclear Operating Corporation, Kansas Gas and Electric Company, Kansas City Power &Light Company, Kansas Electric Power Cooperative, Inc., Docket No. 50-482, WolfCreek Generating Station, Coffey County, Kansas Date ofamendment request: July 26,

'1988 Briefdescription ofomendment request: The proposed amendment would revise Technical Specification 5.3.1, Fuel Assemblies, to allow the-

'eplacement of a limited number of fuel rods with fillerrode or vacancies ifsuch replacement is acceptable based on the results of a cycle-specific reload analysis.

Date ofpublication ofindividual nolicein Federal Register September 9, 1988 (53 FR 35136).

Expiration dole ofindividualnotice:

October 11, 1988 LocolPublic Document Room Location: Emporia State University, WilliamAllen White Library, 1200 Commercial Street, Emporia, Kansas 66801 and Washburn University School of Law Library, Topeka, Kansas NOTICE OF ISSUANCE OF AMENDiyIENTTO FACIUTY.

OPERATING LICENSE 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 and Proposed No Significant Hazards Consideration Determination and Opportunity for Hearing in connection with these actions was published in the Federal Register as indicated. No request for a hearing or petition for leave to intervene was filed followingthis notice.

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. Ifthe Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.12(b) and has made a determination based on that assessment, it is so indicated.

For further details with respect to the action see (1) the applications for amendments, (2) the amendments, and (3) the Commission's related letters, Safety Evaluations and/or Environmental Assessments as indicated. Allof 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. A copy of items (2) and (3) may be obtained upon request addressed to the U.S. Nuclear Regula tory Commission. Washington.

DC 20555. Attention: Director. Division ofReactor Projects.

Consumers Power Company, Docket No.

50-1 55, Big Rock Point Plant, Charlevoix County, Michigan Date ofopplicotion foramendment:

December 2, 1986. and February 1, 1988.

Briefdescription ofamendment: This amendment modifies paragraph 2.C.(5) of the license to require compliance with the amended Physical Security Plan, This Plan was amended to conform the requirements of 10 CFR 73.55. Consistent with the provisions of 10 CFR 73.55, search requirements must be implemented within 60 days and miscellaneous amendments within 180 days from the effective date of this amendment.

Date ofissuonce: September 14, 1988 Effecti ve date: September 14, 1988 Amendment No/t 92 Facility Operating License No. DPR-6, The amendment revised the license.

Date ofinitiolnoticein Federal Register. May 4, 1988 (53 FR 15909). The Commission's related evaluation of the amendment is contained in a letter to Consumers Power Company dated September 14, 1988 and a Safeguards Evaluation Report dated September 14, 1988.

No significant hazards consideration comments recei ved: No.

Local Public Document Room location: North Central Michigan College, 1515 Howard Street, Petoskey, Michigan 49770.

Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire NucIear Station, Units 1 and 2, Mecklenburg County, North Carolina Date ofopplication foromendments:

June 24, 1988 Briefdescription ofamendments: The amendments modified the Technical Specifications by deleting surveillance requirements regarding manual transfer from normal to emergency power supplies for the pressurizer heaters, the power-operated relief valves (PORVs),

and the PORV block valves.

Date ofissuonce: September 13, 1988 Effective date: September 13, 1988 Amendment Nos.: 92 and 73 Facility Operating License Nos. NPF-9 and NPF-17: Amendments revised the Technical Specifications.

Federal

/ VoL 53. No. 193 f Wedaesdiy.

Oc 5, 1888 '/ %Nues Dale ofinitMIaolicein Federal Register. hugest 10, 2988 ($3 FR X630).

The Commission's rdated evaiuahon of the amendments is contained in a Safety Evaluation dated Septeinber13, 198L No significant hazatris consideration comments recei ved: No.

Local Public Dacumenl Room location: Atkins Library, University of North Carolina. Charlotte (UNCC

" Station);North Carolina 28223 Florida Power and tight Company, et al.,

Docket No. 50-389. St. Lucie Plant, Unit No. 2, St. Lucle County, Florida Date ofapplicalian foramendment:

January 25, 1985 (partial)

Briefdescription ofameadmeaL The amendment deleted various license conditions as well as Attachment 1.

Appendix E, and Appendix F to the License.

Date ofIssuance: September 13, 1988 Effective Dale: September 13, 1988 Amendment No '34 Facility Operating License No. NPF-16t Amendment revised the License.

Date ofinitialnoticein Federal Register: May 21. 1985 (50 FR 20976). 'IIie Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 13, 198L No significant hazards consideration comments recei vedi No.

Local Public Document Room localiotu Indian River Junior College Library, 3209 Virgina Avenue, Ft. Pierce, Florida.

Florida Power and Light Company, et al Docket No. 50-389, St. Lucie Plarit, Unit No. 2, St. Lucie County, Horida Dale ofapplication ofamendmenr November 1L 1987 Briefdescriplian ofamendment: The amendment revised Sections 4.7.24 and 4.7.1.6 of the Technical Specifications for the main steam isolation valves and the main feedwater isolation valves.

respectively.

Date ofIssuance: September 22. 1988 Effective Date: September 22, 1966 Amendment No.: 35

~

Facility Operaling License ¹. NPF-16: Amendment revised the Technical Specifications.

Date ofinitialnoticein Federal Register: December 30, 1987 (52 FR 49227). The Commission's rated evaluation of the amendment is contained in a Safety Evaluation dated September 22, 1988.

Na significant hazards consideration comments recei ver No.

Local Public Document Roonr location: Indian RiverJ~ College Library.3209 Virgina Avenue, Ft. Pierce, Florida.

Georgia Power Gempany, Oglethorpo Po'iver Corporatio< hglmicipal EIactric AuthorityofGeorgia, CityofDalton, Georgia, Docket Nos.56-822 end 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2, Appling County, Georgia Dale ofapplicatioa far amendmeats:

June 20, 1988 Briefdescriptioa afamendmenls: The amendments modify the Technical Specifications to delete all references to the main control room chlorine detectors and to the automatic isolation of the main control room environmental control system on high chlorine level.

Date ofissaance September 12. 198S Effec&e date.'eptember 22, 1988 Amendment Nos.: 156 and 96 Facility Operating License Nos. DPR-57 and NPF-5, Amendments revised the Technical Specifications.

Dole ofinitialnoticein Fedeial Register. August 10, 1988 (53 FR 30135),

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 12, 198L No significonl hozards consideration commeals received: No.

Loca/Public Document Roam location: Appling County Public Library, 301 City Hall Drive. Baxley. Georgia 31513 Georgie Power Company, Oglethorpe Power Corporation. MunicipalElectric AuthorityofGeorgia,'ity ofDalton, Georgia, Docket No. 50-321, Edwin L Batch Nuclear Plant, Unit1,Aping County, Georgia Date ofapplication foramendment:

June 20, 2986 Briefdescripton ofamendment: The amendment modified the Technical Specificathms to allow lhe use of General Electric SxSKB fueland lead fuel assemblies produoed by Advanced Nuclear Fuels.

Dale ofissuance: September 22, 1986 Effective date September 22, 1988 Amendmeul No.: 157 Faci%'ty Opemting License hb. DPR-

57. The amendment revised the Technical Specifications.

Dale ofinitialnoticein Federal Register: August 10, 1988 (53 FR 30232).

The Commission's related evaiuation of the amendment is contaiuiul in a Safety Evaluation dated September 22, 2988.

No sjgnificant hazards consideration comments received hh.

Local PabL'c Dacamaut Roaur location: Appling Csuuity.Pubhc Library,,

301 City HallDrhwBaxley. Georgia

'1513 GPU Nuclear Corporations at irI., Docket No.56-219, Oyster Creek Nuciaar Generating Station, Ocean Cmmty, New Jersey Date ofapp!icatian forameadnierit:

October 23, 1988 as supplemented April 5 1988.

Briefdescription ofamendment: The amendment revised Technical Specifications (TS) to reflect changes in the requirements on the maximum radioiodine concentration allowed in the reactor coolant in TS Sections 3.8 and 4.6.

Dale ofIssuance: September 12. 1988 Effective date: September 12. 198S Amendment No.: 126 Provisional Operating License No.

DPR-18. hmendment revised the Technical Specifications.

Dote ofinilialnoticein Federal Register. May 18. 1988 (53 FR 17789). The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated September 12, 1988.

No significant hazards consideration cammenls recei ved: YeLBy letters dated November 7, 1986 and December 31, 2986, the Bureau of Engineering.

Division ofEnvironmental Qualify.

Department ofEnvironmental Protection. State ofNew Jersey raised concerns. By letter dated July 20, 2987, the staff responded to the conceriui. No other coruments were receiveL Local Public Document Roam location: Ocean County Library.

Reference Department, 101 Washington Street. Toms River, New Jersey 08753.

Mississippi Pawl it Light Ccaayauy.

System Energy Resources. Inr South Mississippi Eloctcic Power Associatiaa, Docket No. 50-416, Grand GulfNuclear Station, Unit 2, Claiborne County, Mississippi Dote ofapplicatioa foramendment July 15, 1988 Briefdescription ofamenrimeaL The amendment would change the Technical Specifications by changing the position of one of the nine members on the Plant Safety Review Committee (PSRC) fiaru Technical Support Superintendent to Manager, Performance and System Engineering. This change is necessitated by a change of the unit organizahan to consolidate certain enginoermg personnel and functions.

Date ofissuance September 2L 1988 Effective dote'eptember 22, 19tS Amendment Na 47 FacilityOperatr'ng Licerrsel&. NPIr-

29. Thhi amendment revises the Technical Speoificationa Dale ofinitialnetice in Federal Register. August 10, 1968 (53 FR'30139).

39182 Federal Register oi. 88, No. 198 / Wednesday.

Octoberl988

/ Notices The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 21, 1988.

No significont hazards considerotion comments recei ved: No Local Public Document Room location: Hinds Junior College, McLendon Library, Raymond, Mississippi 39154 Mississippi Power & Light Company, System Energy Resources, Inc., South Mississippi Electric Power Association, Docket No. 50-416, Grand GulfNuclear

"'tation, Unit 1, Claiborne County, Mssissippi Date ofopplicotion foramendment:

July 12. 1988

\\

Briefdescription ofamendment: The amendment deletes Room OC506 from Table 3.3.7.9.1 as a result of a design change to enlarge the control building locker room.

Date ofissuance: September 23, 1988 Effective date: September 23, 1988 Amendment No. 48 Facility Operoling License No. NPF-

29. This amendment revises the Technical Specifications.

Date ofinitialnoticein Federal Register. August 10, 1988 (53 FR 30138).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 23, 1988.

No significont hazards consideration comments recei ved: No Local Public Document Room location: Hinds Junior College, McLendon Library. Raymond, Mississippi 39154 Northeast Nudear Energy Company, et al., Docket No. 50-423, Millstone Nudear Power Station, Unit No. 3, New London County, Connecticut Date ofopplication foramendment:

February 24. 1988 Briefdescription ofamendment: The amendment changes Technical Specification (TS) 3.3.3.9, "Radioactive Liquid Effluent Monitoring Instrumentation" and TS 3.3.3,10, "Radioactive Gaseous Monitoring Instrumentation." The changes provide for the following: (1) allowance for planned inoperability of monitoring-instrumentation for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for the purpose of maintenance and performance of required tests, checks, calibration or sampling. (2) a requirement to initiate auxiliary sampling within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after inoperability of certain gaseous effluent monitors, and (3) allowance for inoperability of certain liquid effluent monitoring instrumentation. during Mode 6 (refueling), when the effluent pathway is not being use.

Dote ofissuance: September 9, 1988 Effective date: September 9, 1988 Amendment No.: 22 Focility Operating License No. NPF-

49. Amendment revised the Technical

'pecifications.

Dote ofinitialnoticein Federal Register. July 27, 1988 (53 FR 28292). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 9, 1988.

No significont hozords consideration comments recei ved: No.

Local Public Document Room location: Waterford Public Library, 49 Rope Ferry Road, Waterford, Connecticut 06385.

Public Service Company of Colorado, Docket no, 50-267, Fort St. Vrain Nudear Generating Station, Platteville, Colorado Date ofomendment request: February 5, 1988, as supplemented June 23, 1988.

Briefdescription ofamendment: This amendment changed certain portions of the Administrative Controls section of the Technical Specifications. The portions concern the licensee's organization and the Plant Operations Review Committee.

Date ofissuonce: September 15, 1988 Effective date: September 15, 1988 Amendment No.: 83 Facility Operating License No. DPR-

34. Amendment revised the Technical'peciTications.

Date ofinitiolnotice in Federal Register. August 10, 1988 (53 FR 30142).-

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 15, 19M.

No significant hazards consideration comments received: No.

LocolPublic Document. Room location: Greeley Public Library, City Complex Building, Greeley, Colorado Public Service Company of Colorado, Docket No. 50-287, Fort St. Vrain Nudear Generating Station, Platteville, Colorado Date ofamendment request: April20.

1988. and supplemented July 1 and August 5, 1988.

Briefdescription ofamendment. This amendment made certain changes to the Technical Speciflicatfons for the plant's DC power systems. It also allowed for future changes to the station batteries.

Date ofissuance: September 15. 1988 Effective dote: September 15, 1988 Amendment No,'4 Facility Operating License No. DPR-34 Amendment revised the Technical Specifications/license.

Dote ofinitialnoticein Federal Register. June 15, 1988 (53 FR 22405). The licensee'.s. July 1. 1988 submittal provided reformatted pages for section 4.6 of the Technical Specifications. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 15:1988.

No significont hazards consideration comments received/ No.

LocolPublic Document Room location: Greeley Public Library. City.

Complex Building, Greeley, Colorado Tennessee Valley Authority, Docket Nos. 50-S27 and 50-328, Sequoyah Nudear Plant, Units 1 and 2, Hamilton County, Tennessee Date ofapplication foramendments:

June 13, 1988 (TS 88-01),

Briefdescription ofomendments: This amendment revises Table 3.6-2, "Containment Isolation Valves." of the Sequoyah Units 1 and 2 Technical Speciflication (TS). The revisions are to add five motor-operated valves (MOVs) to the table. These MOVs are replacing check valves as containment isolation valves. The amendment also adds a note to Table 3.3-5, "Engineered Safety Features Response Times." to reflect that the response times of these MOVs, when they are actuated by a Phase B containment isolation signal, are slightly longer than other containment isolation valves.

Dote ofissuance: September 9, 1988 Effective dote: September 9, 1988 Amendment Nos, 82, 73 Facility Operating Licenses. Nos.

. DPR-77 and DPR-79. Amendments revised the Technical Spedfications.

Date ofinitialnoticein Federal Register. July 13, 1988 (53 FR 26533). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 9, 1988.

No significant hozards consideration comments recei vedi No Local Public Document Room location: Chattanooga-Hamilton County Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

Tennessee Valley Authority, Docket Nos. I-327 and 50-S28, Sequoyah Nudear Plant, Units 1 and 2, Hamflton County, Tennessee Date ofopplicotion for amendments:

June 10, 1987 (TS 87-35)

Briefdescription ofamendments:

These amendments revise Table 3.4-1, Reactor Coolant System Pressure Isolation Valves, of the Sequoyah. Units 1 and 2 Technical SpeciTications (TS).

The changes are to add the two upper head injection charging header valves to Table 3.4-1. These valves are different from most of the valves in Table 3.4-1 in that they do not have to be leak tested followingmanual or automatic actuation or flow through the valve. In its

N Federal-Register / Vol.. 53, No. 193 / Wednesday.

October 5, 1988</ Notices, 39183 application, the Tennessee Valley Authority (TVA)also withdrew its TS change 68 which it had submitted in its letter dated May 10, 1986.

Dole ofissuance: September 21, 1988 Effective date: September 21, 1988 Amendment Nos.t 83. 74 FocilityOperating Licenses Nos.

DPR-77 and DPR-79. Amendments revised the Technical Specifications.

Dale ofinitiolnotice in Federal Register. November 4, 1987 (52 FR 42370.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 21. 1988.

No significont hazards consideration comments recei vedt No Locol Public Document Room locoliont Chattanooga-Hamilton County Library. 1001 Broad Street, Chattanooga.

Tennessee 37402.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee Dote ofopplicalion for amendmentst February 27, 1987 (TS 82)

Briefdescription ofamendments:

These amendments revise Specification 3/4.4.1.2, Reactor Coolant System, Hot Standby. in the Sequoyah Units 1 and 2 Technical Specifications (TS). The changes are to increase the number of reactor coolant system loops required to be in operation during Mode 3, Hot Standby. to two loops. The TS limiting condition for operation, action statement and surveillance requirement are being revised. The Bases for the Specification 3/4.4.1.2 are also being changed.

Dole ofissuance: September 22, 1988 Effective dote: September 22, 1988 Amendment Nos.t 84, 75 FacililyOperating Licenses Nos.

DPR-77 and DPR-79. Amendments revised the Technical Specifications.

Date ofinitialnoticein Federal Register. August 12, 1987 (52 FR 29928).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 22. 1988.

No significant hazards consideration comments received: No Local Public Document Room locotiont Chattanooga-Hamilton County Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

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

May 22, 1987 (TS 87-18)

Briefdescription ofamendments:

These amendments revise the reactor trip limits for reactor coolant pump undervoltage in Table 2.2-1, Reactor Trip System Instrumentation Trip Setpoints, of the Sequoyah Units 1 and

?.Technical Specifications (TS). The minimum reactor trip setpoint is being increased for each bus from 4830 volts to 5022 volts. The minimum allowable values is being decreased for each bus from 4761 volts to 4739 volts.

Dale ofissuance: September 22, 1988 Effective date: September 22, 1988 Amendment Nos, 85, 76 Facility Operating Licenses Nos.

DPR-77 ond DPR-79. Amendments revised the Technical Specifications.

Dale ofinitialnoticein Federal Register. August 12, 1987 (52 FR 29933) ~

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 22, 1988.

No significant hazards consideration comments received: No LocolPublic Document Room localiont Chattanooga-Hamilton County Library. 1001 Broad Street, Chattanooga, Tennessee 37402.

Toledo Edison Company and The Cleveland Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear Power Station, Unit No. 1, Ottawa County, Ohio Date ofopplicati on foromendmentt December 8, 1987 Briefdescription ofamendment: The amendment revised TS 4.6.1.2.c.3 to be consistent with the requirements of Appendix J to 10 CFR Part 50 and ANSI N45.4-1972 Appendix C.

Dote ofissuance: September 19, 1988 Effective dote: September 19, 1988 Amendmenl No. 120 Facility Operating, License No. NPF-3.

Amendment revised the Technical SpeciTications.

Date ofinitiolnolicein Federal Register. April6. 1988 (53 FR 11378). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 19. 1988.

No significant hazards consideration comments receivedt No Local Public Document Room location: University ofToledo Library, Documents Department. 2801 Bancroft Avenue, Toledo, Ohio 43606.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, Vermont Dote ofapplication foramendment:

May 23, 1988, as supplemented on August 15, 1988.

Briefdescription ofamendment: The amendment revises the Technical Specifications to permit the use of the fuel type designated as GE BXBEB.

,, Date ofissuance: September 9. 1988 Effective date: 30'days from'date of.

issuance Amendment No,'108 Facility Operating License iVo. DPR-ZSt Amendment revised the Technical SpeciTications..

Date ofinilialnoticein Federal Register. June 15, 1988 (53 FR 22408). The Commission's related evaluation of the amendment is contained in 8 Safety Evaluation dated September 9, 1988.

No significant hazards consideration comments recei vedi No.

Local Public Document Room Locotiont Brooks Memorial Library, 224 Main Street. Brattleboro. Vermont 05301.

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee Nuclear Power Plant, Kewaunee County, Wisconsin Dole ofopplication for amendment:

October 26. 1987, supplemented June 16, 1988 Briefdescription ofamendment: The amendment revised the Technical Specifications (TS) to clarify existing specifications and increase the consistency within the TS.

Date ofissuonce: September 20, 1988 Effective date: September 20, 1988 Amendment No.t 80 Faci%'ty Operoting License No. NPF-

30. Amendment revised the Technical Specifications.

Date ofinilialnoticein Federal Register. July 13, 1988 (53 FR 26536). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 20. 1988.

No significant hazotds consideration commenls received: No.

Loco/ Public Document Room locotiont University of Wisconsin Library Learning Center. 2420 Nicolet Drive, Green Bay, Wisconsin 54301.

Dated at Rockvllle, Maryland. this 27th day of September. 1988.

For the Nuclear Regulatory Commission

. Steven A. Varga, Director. DivisionofReactor Prat'ects-I/IL OfficeafNuclear Reactor Regulation (FR Doc. 88-22808 Filed 10-4-88: 8:45 am) stLLINo coos 7$~w)

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