ML16341E358
| ML16341E358 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 07/17/1987 |
| From: | Kirsch D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | Shiffer J PACIFIC GAS & ELECTRIC CO. |
| References | |
| NUDOCS 8707290229 | |
| Download: ML16341E358 (38) | |
Text
ACCESSION NBR FAC IL: 50-275 50-323 AUTH. NAME KIRSCH> D. F.
RECIP. NAME SHIFFER> J. D.
REGULA Y INFORMATION DISTRIBUTI
'.SYSTEM (RIDS) 8707290229 DOC. DATE: 87/07/i7 NOTARIZED:
NO DOCKET Diablo Canyon Nuclear Poeer Plant> Unit'> PaciFic Ga 05000275 Diab lo Canyon Nuclear Power Plant>
Unit 2.
Pac ific Ga 05000323 AUTHOR AFFILIATION Region 5>
OFfice of Director RECIPIENT AFFILIATION PaciFic Gas 8< Electr ic Co.
SUBJECT:
Conf irms 8707i5 telcon e/MM Mendonca re mgt meeting to be held beteeen NRC
- 8. util representative in conference room at i. QO
- p. m.
on 870817.
DISTRIBUTION CODE:
IEOiD COPIES RECEIVED: LTR J ENCL Q SIZE:
TITLE: General (50 Dkt)-Insp Rept/Notice of Violation Response NOTES:
RECIPIENT ID CODE/NAME PD5 PD INTERNAL:
ACRS DEDRO NPR/DOEA DIR NRR/DREP/RPB NRR/PMAS/ ILRB OGC/HDS2 RES DEPY GI EXTERNAL:
LPDR NSIC COPIES LTTR ENCL 2ii ii RECIPIENT ID CODE/NAME TR*MMELL,C AEQD NRR MOR ISSEAU> D NRR/DREP/EPB NRR/DRIS DIR MAN> J RGN5 FILE Oi NRC PDR CQP IES LTTR ENCL TOTAL NUMBER OF COPIES REQUIRED:
LTTR 23 ENCL RS
July 17, 1987 Docket Nos.
50-275 and 50-323 Pacific Gas and Electric Company 77 6eale Street, Room 1451 San Francisco,'alifornia 94106 Attention:
Mr. J.
D. Shiffer, Vice President Nuclear Power Generation Gentlemen:
Subject:
Significant Meeting This will confirm your telephone discussion with Mr. M. M. Mendonca of this office on July 15, 1987 concerning a management meeting to be held between NRC and Pacific Gas and Electric Company representatives in the Region V
conference room at 1:00 p.m.
on August 17, 1987.
This meeting is to reschedule our July 16, 1987 meeting, in accordance with your request to give you and your management the opportunity to be more directly involved in the current Unit 2 restart activities.
Also included is a copy of our internal meeting notice which lists the anticipated attendees at this meeting.
Sincerely,
/g/
Enclosures:
Notice of Significant Meeting.
Dennis F. Kirsch, Director Division of Reactor Safety and Projects RV/jk C
YES /
NO YES./
NO MENDONCA CHAFFEE 7/(g/87 I/jr(/87 YES NO T
PDR YES /
NO KIRSCH 7/y /87 8707290226 870717 PDR ADOCK 05000275 P
i 4.
RETSSUED PITH ENCLOSURE
'DI'STRIBUTION July 15, 1987 LPDR PD5 Rdg JLee (2)
DOCKET NO(S).
50-275/323 Mr. J.
D. Shiffer, Uice President Nuclear Power Generation c/o Nuclear Power Generation, Licensing P acific Gas and Electric Company 7 7 B cele Street, Room 1451 S sn Prancisco, CA 94106
SUBJECT:
PACIPIC GAS AND ELECTRIC COMPANY DIABLO CANYON NUCLEAR POWER PLANT
\\
The following documents concerning our review of the subject facility are transmitted for your information.
Notice of Receipt of Application, 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 g) Bi-Weekly Notice; Applications and Amendments to Operating Licenses Involving No Significant Hazards Considerations, dated 7/1/87 Lace page(s) ]
24565 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 dated Annual/Semi-Annual Report-transmitted by letter dated Encl osures:
As stated Office of Nuclear Reactor Regulation See next page OFFICE/
SURNAME)
OA~E f>
DRS lPD 7/1 87
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NRC FORM 3I8 IIO/SOI NRCM 0240
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OFFICIAL RECORD COPY
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Hr. J.
D. Shiffer Pacific Gas and Electric Company Diablo Canyon CC:
Richard F. Locke, Esq.
Pacific Gas
& Electric Company Post Office Box 7442 San Francisco, California 94120 Janice E. Kerr, Esq.
California Pub11c Ut11ities Commission 350 McAllister Street San Francisco, Cal1fornia 94102 Ms. Sandra A. Silver 660 Granite Creek Road Santa Cruz, California 95065 Mr. W. C. Gangloff Westinghouse Electric Corporation P. 0. Box 355 Pittsburgh, Pennsylvania 15230 Hanaging Ed1tor San Lu1s Obispo County Telegram Tribune 1321 Johnson Avenue P. 0.
Box 112 San Luis Obispo, Californ1a 93406 Q. Leland M. Gustafson, Manager Federal Relations Pacif1c Gas and Electric Company 1726 H Street, N.
'W.
Washington, DC 20036-4502 Dian M. Grueneich, Esq.
Edwin F. Lowry, Esq.
Grueneich
& Lowry 380 Hayes Street Suite 4
San Francisco, Cal1fornia 94102 NRC Resident Inspector Diablo Canyon Nuclear Power Plant c/o U.S. Nuclear Regulatory Comm1ss1on P. 0. Box 369 Avila Beach, Cal1forn1a 93424 Mr. Dick Blakenburg Ed1tor & Co-Publisher South County Publishing Company P. 0. Box 460 Arroyo Grande, Cal1forn1a 93420 Bruce Norton, Esq.
c/o Richard F. Locke, Esq.
Pacific Gas and Electric Company Post Office Box 7442 San Francisco, Californ1a 94120 Dr. R. B. Ferguson Sierra Club - Santa Luc1a Chapter Rocky Canyon Star Route Creston, California
. 93432 Chairman San Luis Obispo County Board of Supervisors Room 220 County Courthouse Annex San Luis Obispo, Californ1a 93401 Director Energy Faci 1it1es S1ting Division Energy Resources Conservation and Development Comm1 ss1on 1516 9th Street Sacramento, Cal if orn1a 95814 Hs. Jacquelyn Wheeler 2455 Leona Street San Luis Obispo, California 93400
I
=- Pacfffc Gas 5 Electrfc=:Company', "-..'-
0
'iablo Canyon'-
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Ms. Laurie McDermott, Coordinator Consumers Organized for Defense of Environmental Safety 731 Packffc Street, Suite 42 San Luis Obispo, California 93401 Mr. Joseph
- 0. Ward, Chief Radiological Health Br anch State Department of Health Services 714 P Street, Office Building 0'8 Sacramento, Ca 1 ffornia 95814 Regional Administrator, Region V
U.S. Nuclear Regulatory Commission 1450 Maria Lane Suite 210 Alnut Creek, California 94596 Ms. Nancy Culver 192 Luneta Street San Luis Obispo, Calffornfa 93401 President Calffornfa Public Utilities Commission California State Building 350 McAllfster Street San Francisco, California 94102 Michael M. Strumwasser, Esq.
Special Assistant Attorney General State of California Department of Justice 3580 Wflshire Boulevard, Room 800 Los Angeles, California 90010
24642
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Federal Register / Vol." S2, No. 128 / Wednesday, July 1. 19M' Notices NUCLEARREGULATORY COMMISSION
(&weeklyNotice AppIIcatlona and Amendmenta to Operating Licenses InvoivlnIINo Significant Hazards L Background Pursuant to Public Law (P.L) 97415, the Nuclear Regulatory Commission (the Commission) is publishing this regular bi-weekly notice. P.L 97<15 revised section 189 of the Atomic Energy Act of 1954. as amended (the Act), to require the Commission to publish notice of any amendments issued. or proposed to be issued. under a new provision of section 189 of the Act.'Ibis provision grants the Commission the authority to issue and make immediately effective any amendment to an operating license upon a determination by the Commission that such amendment involves no significant hazards consideration. notwithstanding the pendency before the Commission of a request for a hearing from any person.
This bi-weekly notice includes all notices of amendments issued, or proposed tttbe issued from June 8, 1987 through June 19, 1987. The last bi-weekly notice was published on June 17, 1987 (52 FR 23092).
NOTICE OF CONSIDERATIONOF ISSUANCE OF AMENDMENTTO FACILITYOPERATING LICENSE AND PROPOSED NO SIGNIHCANT HAKQtDSCONSIDERATION DETHMINATIONAND OPPORTUNITY FOR HEARING The Commission has made a proposed determination that the following amendment requests Involve no significant hazards consideration. Under the Commission's regulations in 10 CFR 50.92, this means that operation of the facilityin accordance with the proposed amendments 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
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1 Federal Regjster
/., Vol. 62, No. 128 / Wednesday, July 1, 1987 / Notices any accident previously evaluated; or (3)
Involve a significan reduction In a margin of safety. The basis for this proposed determinafion for each amendment request Is shown below.
The Commission Iayeeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice willbe considered In making any final determination. The Commission willnot normally make a final determination unless it receives a request for a hearin j.
Writt'en coinments may be submitted by inail to the Rules and Procedures Branch, DivisionofRules and Records, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and should cite the publication date and page number of tlils Federal Reghiter notice. Written comments may also be delivered to Room 4000, Maryland National Bank Building, 7735 Old Georgetown Road, Bethesda, Maryland from 8:15 a.m. to 5:00 p.m. Copies ofwritten comments received may be examined at the NRC Public Document Room, 1717 H Street, NW, Washington. DC. The filingof requests for hearing and petitions for leave to intervene is discussed below.
By July 31, 1987. the licensee may file a request for a hearing with respect to issuance of the 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 the proceeding must file a written petition for leave to intervene. Requests for a hearing and petitions for leave to intervene shell be filed'in accordance with the Commission's "Rules of Practice for Domestic Licensing Proceedings" in 10 CFR Part 2. Ifa request for a hearing or petition for leave to intervene is filed by the above date. the Commission or an Atomic Safety and Licensing Board. designated by the Commission or by the Chairman of the Atomic Safety and Licensing Board Panel. willrule on the request and/or petition and the Secretary or the designated Atomic Safety and Licensing Board willissue a notice ofhearing or an appropriate order.
As required by 10 CFR 2.714. a petition for leave to intervene shall set fcrth with particularity the interest of t!ie petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following factors: (1) the nature of the petitioner's right under the Act to be made a party to the proceeding, (2) the nature and extent of the petitioner's property, financial, or other interest In the proceeding, and (3) the possible effect of any order which may be'ntered in the proceeding on the petitioner's interest. The petition should also Identify the specific aspect(s) of the subject matter of the proceeding'as to which petitioner wishes to Intervene.
Any person who has filed a petition for leave to int'ervene or who has been admitted as a party may amend the petition without requesting leave of the Board up to fifteen (15) days prior to the first prehearing conference scheduled in the proceeding, but such an amended petition must satisfy the specificity requirements described above.
Not later than 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 crosswxamine witnesses.
Ifa hearing is requested, the Canlliilssion willmake a final determination on the issue of no 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 SMay notice period.
However. should circumstances change during the notice period such that failure to act in a timely way would result, for I exemple, in derating or shutdown ofthe facility, the Commission may issue the license amendment before the expiration ofthe 30-day notice period,
, provided that its final determination Is th'at the amendment involves no, ',
significant hazards consideration. The final detenninatlon willconsider all public and State comments received before action is taken. Should the Commission take this action, it will' publish a notice of issuance and provide for opportunity for a hearing after issuance, The Commission expects that the need to take this action willoccur very inf'requently..
Arequest for a hearing or a petition for leave to intervene must be filed with the Secretary of the Commission, U,S.
Nuclear Regula tory Commission, Washington, DC 20555, Attention:
Docketing and Service Branch, or may, be deliv'ared to &e Commission's Public Document Room, 1717 H 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 (800) 3~~
(in Missouri (800) 3424700).
The Western Union operator should be given Datagram Idcntification Number 3737 and the foHawlng message addressed to (Pleat Director):
petitioner's name and telephone number, date petition was mailed; plant name; and publication date and page number of this Federal Register notice.
A copy of the petition should also be sent to the Office of the General Counsel-Bethesda, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and to the attorney for the,
'icensee.
Nontimely filings ofpetitions for leave to intervene, amended petitions, supplemental petitions and/or requests for hearing willnot be entertained absent a determination by the Commission. the presiding ofHcer 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)-(v) and 2.714(d).
For further details with respect <o this action. see the application for amendment which Is available for public inspection at the Commission's Public Document Room. 1717 H Street, NW, Washington, DC, and at the local public document room for the particular facility, invalved.
0 Federal Regear / Va). S2, No. 126 / Wedttesday, July I, XNF -( Ãa6ces
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Arkansas Power 6i LIght Ceealtaci5t, Docket No. 64313;Atkaasas Nachar One, Unit 1, Pope County, Aticansm Date ofamendment.request:
December 12, 1966 Description ofamendntaat reqiiest The proposed amendment would: (1) change the maximum cote eatichtaent specification in Technical Specification (TS) Section 5~4 from ".shaH not'xceed an enrichment'of 3.6 percent of
'235v" to "...shall be of law enrichment.,
and (2) change the active height value of the core in TS 52.12 &om "M44 inches" to "...approximately 142 inches."
Basis forproposed no significant hazards consideration determinatian:
The inclusion of the maximum enrichment of the fuel is not a dliect input to the reactor safety analysis. The fuel enrichment is used in an indirect manner in conjunction with a number of other parameters in performing the nuclear design of the reactor fuel cycle.
The nuclear design is used in ban to derive measurabki cora parameters important to safe operation which are included in the TS as Lhnititrg Conditions for Operation. Ia adcithn.
the reload fuel enridanent is Included in the fuel storage TS and gives the maximum enrichment ofaew fuel which can be stored in the spent fuel pooL The active height of the cora varies slightly from reload to rehiacL Core heights are Identified and evaluated as part of the safety analysis for each cora The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 5082. Aproposed amendment to an operating Hcense for a facilityinvolves no significant hazards consideration Ifoperation of the facIBty in accordance with the proposed amendments would not: (1) Involve a signiTicant increase in the probabIHty or consequences of an accident previously evaluatech or (2) Create the possibIHty of a new or different kind of accident fmm any accident previously evaluated; or (3)
Involve a significant reduction in a margin of safety.
A discussion of these standards as they relate to the proposed change follows:
(I)Consideration ofProbability and Consequences ofAccident: The fuel enrichment is included along with other factors in performing the nuclear design of fuel which in'urn is subjected to a safety analysis prior to reload ln accordance with NRC methodology. The speciffcation of fuel enrichment in the core design section above does not uniquely determine the values ofthe reactor ccee parameters which ate important to safety. Small changes is core height (2:to 3 inches out oT 142 inches) are IdentiBed <<nd evaluated as part af the core rehed report. Becaese chaages in the cate are already Included in a safety analysh and must be In accordance with speciSc criteria, the proposed change willnot involve a signthant,increase in the probabIBty or cansequences of an accident previously evaluated.
(2) Cant<< the Probabilityofa¹tvor DifferentAccident: As noted eetBar, the enrichiaent aad care.ltelght ate f<<ctors whIch are aheady Included in the.
derivation of case parameters which are included in the Technical Speciffc<<Hone.
Therefare the pmposed changes willnot create the possibiHty af a new or different kind ofaccident from any accident previously evahiated.
(3) Involvea Significant Beductianin a MarginofSafety: Fue) enrichment and core height are considered during the core reload calculations and in the derivation ofmeasurable core
'arameters important to safe operation.
The core parameters hnportaat to safety must be within speciBed Bmits as stated in the Technical SpeciffcationL Tlien.
the proposed changes willnot htvolve a significant reduction in the margin of safety.
Based oa the above coastdetatians, the Commission proposes to deteradne that the propeied changes'Involve ao significant hazards cansidera6oaL Local PubDc Document Boom location: Tomlinsan LIbraty, Arkansas Tech University, Russellville. Arkansas 72801 Attorneyforlicensea Nicholas S.
Reynolds, EsqBishop, Liberman, Cook.
PurceH and Reynolds, 1200 Seventeenth Street, NW, Washington. DC 20036 NRCeject D rector. Jose A. Calvo Commonwealth RcBsoa Company, Docket Nos. STN 5M64 and STN 50-455, Byren St<<See, Unit Nes. 1 and 2.
Ogh County, IIHnebr, end Docket Na STN SM56 Braidwood Station, UnIt No.
1 WiQ Cocatty, IIHaois Date ofapplication foramendments:
May 20, 1987 Description ofamendments request:
The amendment would revise Technical Specification Tables 43-1 and 4.3-2 to eliminate setpoint verIQcatlon when performing the monthly and quarterly Trip Activating Device Operational Teat for the undervaltage and underfrequency relays. The current Technical Speciffcations contain a nota (Note 3 in Table 43-2) that indicates that setpoint verlffcatian h not required during the Maathly'Mp Actuating Device Operational Test far the GrId Degraded VoBaga The intent ofthh note was that setpoiat verificatio was not required for any ofthe undervoltaga and underfreqaeacy relays more often than every eighteen months (during an outage). It h,tha stafFs Intention to apply this amendmeat, Kitis found acceptable, to Braidwood Station Unit 2,
- when it receives Ih operaUng Bcense.
Sbsis forProposedha Significant Hazards Considnalion Deterntmaticm:
The staffhas evaluated this prapased amendment and has determined that it involves no slgniffcaat hazards consideration. AccorcBng to 10 CFR 50.92(c), a proposed amendmeat to an operating license involves no sigaIBcant hazards considerations lfoperathm of the fadBty hiaccordance with the proposed amendment would'nat: (1)
Involve a signlffcant increase in the probibiBty ar ccmsequeaces ofaa accident previausly evaluated; or (2) create the possilRBty af a new or different kind ofaccident fmm any accident previously evaluated: or (3) involve a sfgaiffcant reduction in a matsht ofsafety.
The proposed amendment adds notes to TaMes 4.3-1 and 4&2, to clarify tha reqairemeats ofthe TripActuating Device Operational Test as itpertains to undervoltage and under&equency relays. The proposed amendment more
'expHcitly defines the requirements of the Trip Actaathig Device Operational Test as itpertains to undervaltage and underfrequency relays. The relay operabIBty veriffcatian remains unaffected. The amount af time that the plant is in a degraded condition would be increaied Ifsetpaiat yeriffcation was done monthly. Therefore, thh does not increase the probabiBty or consequences ofan accident previously evaluated.
The proposed amendment does not involve any hardware changes. The type and frequency of the surveillance remains unchanged. FSAR analyses and system design envelope the loss of Engineered Safety Feature, or Reactor Protection System Functions. Therefore.
this does not create the possibility of a new or dIIferent kind ofaccident from any accident previously evaluahcL The intent ofthe requirement does not encompass setpoint verification at a frequency greater then eighteen (18) months. The proposed amendment serves to clarify tha understanding of the surveIQeace frequeacy. Therefore, this does not involve a signiffcant reduction in a margin ofsafety.
Therefore, based on the abovct consideratioas.
the staff has determined that these changes involve ao significant hazards conidderatioaa LocaLPublic Docu<<tent Boom locatiotu For Byron Station the Rockford
0 Federal Rag(ster / Vol, 52, No. 126 / Wednesday, July 1," 19Fl / Not(cee, 2454$
Public Library, 215 N. Wyman Street, Rockford, Illinois 81103; for Braldwpod Station the WilmingtonTownship Public Library, 201 S. Kankakee Street,
Wilmington, Illinois60481.
Attorneyforlicensee: Michael Miller,'
Isham, Lincoln 5: Beal, One First National Plaza, 42nd Floor, Chicago, Illinois80%3.
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Connecficut Yankee Atomic Power Company, Docket No. 50-21$, Haddam Neck Plant, Middlesex County, Connecticut Dote ofomendment request: June 1,
'987 Description ofamendment request:
During the 1987 outage, modifications,
'ncluding the installation ofnew motor-operated valves, willbe made to the existing emergency core cooling systems (ECCS) to assure the capability of adequate care cooling over the entire range of pipe breaks. The proposed license amendment willrequire (1) new periodic surveillance requirements to ensure correct, valve position, (2) post-maintenance surveillance requirements for the new throttle valves, and (3) new valve and ECCS system retest requirements followingmodifications to any ECCS subsystem that would,alter ECCS flow characteristics.
Basis forproposed no signi%'cant, hazards consideration determination: In accordance with 10 CFR 50.92. the licensee has reviewed the proposed changes and has concluded that it does not involve a significant hazards consideration. Tlte basis for this conclusion is that the three criteria of 10 CFR 50.92(c) are not compromised: a conclusion which is supported by the licensee's determinations made pursuant to 10 CFR 50.59. The proposed change does not involve a significant hazards consideration because the change would not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated. As stated above. these changes ensure that the present system configuration is maintained, therefore. the probability of occurrence of the design basis accidents is unchanged. Since adequate LOCA mitigation is maintained. the consequences of the design basis accidents are not impacted.
- 2. Create the possibility of a new or different kind of accident from any previously evaluated. There are no new failure modes associated with this proposed change. No new systems or designs are introduced by these proposed changes: therefore. no new failure modes are created. In addition.
I operating characteristics are unchanged.
Thus, no new accident possibilities are created.
- 3. Involve a significant reduction in a margin of safety. As stated above, the proposed changes do not diminish ECCS
. LOCAmitigation capability and thereby
, do not Impact the consequences to the protective bound'aries. Thus, these changes willnot reduce the margin of safety as defined as in any Technical Specifications.
Moreover, the Commission has provided guidance concerning the application of standards set forth in 10 CFR 50.92 by providing certain examples (March 6, 1988, 51 FR 7751) of amendments that are considered not likelyto involve significant hazards consideration. The proposed license amendment is most closely enveloped by example (ii), a change that constitutes an additional control not presently included in the Technical Specifications. The proposed amendment would require new periodic surveillance requirements to ensure valves are in the correct position, new post-maintenance surveillance requirements for the throttle valves, and new system retest requirements followingsignificant modifications to any ECCS subsystem.
Accordingly, the staff proposes to determine that the proposed license amendment does not involve a significant hazards consideration.
Local Public Document Room lacotion: Russell Library, 123 Broad Street. Middletown, Connecticut 08457.
AttotTieyforlicensee: Gerald Garfield, Esquire, Day, Berry and Howard.
Counselors at Law, City Place, Hartford.
Connecticut 08103-3499.
NRCeject Director. Cecil O.
Thomas Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck Plant, Middlesex County, Connecticut Date ofamendment request: June 1, 1987 Description ofamendment request:
The proposed license amendment will revise the heat-up curve (Figure 3.44) to provide a larger interval over which the low-temperature pressurization protection system (LTOPS) must be placed into operation. More specifically, the proposed heat-up rate change results in a pressure/temperature, limitcurve shift. This shift increases the temperature range over which the LTOPS can effectively be placed into operation and decreases the probability of operator error during plant heat-up and cooldown operations.
Basis forproposed no significant hazards consideration detenninatlont In accordance with 10 CFR M.92, the licensee has reviewed the proposed license amendment and has concluded
, that it does not involve a significant hazards consldertttion The basis for this conclusion is that the three criteria of 10 CFR 50.92(c) are not compromised; a conclusion which is supported by the licensee's'eterminations made pursuant to 10 CFR 50.59. The proposed change does not involve a significant hazards consideration because the, change would not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report is not increased since the proposed change results in more restrictive heat.up limitations than presently required while reducing the probability of operator error during LTOPS actuation.
- 2. Create the possibility of a new or different kind of accident from any previously evaluated. The possibility for an, accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report Is not created since the failure to satisfy the requirements of the pressure temperature limitations is similar to a pressurized thermal shock event which has been thoroughly evaluated and documented.
- 3. Involve a significant reduction in a margin of safety. The margin of safety, as defined in the basis for any Technical Specificatio'ns is not reduced since the margin of safety for the present and the new proposed temperature limitcurves are identical.
The staff has reviewed the licensee's determination that the proposed license amendment involves no significant hazards considerations and agrees with the licensee's analyses. Accordingly, the staff proposes to determine that the proposed license amendment does not involve a significant hazards consideration.
Local Public Document Boom location: Russell Library, 123 Broad Street, Middletow, Connecticut 06457.
Attorneyforlicensee: Gerald Garfield, Esquire, Day. Berry and Howard.
Counselors at Law, City Place. Hartford.
Connecticut 08103-3499.
NRC Project Director. Cecil O.
Thomas
Federal Regbter / VoL 5?
No; 12B '/ lN'ednesday, paly 1, %NIL / NoQces ConaacQcut Yankee AtcMafcPower Company, Docket Nc 6041$, Haddara Neck Plant, h6dcBesex County, Connecticut Date ofamendment request: June1, 1987 Description ofamendment request
'IIieproposed Hcense amendment will revise Technical Specfficatfon Section 4.10.D.1 to provide a Iong-term acceptance criteria for the steam generator tubes with defects in the rolled region (bottom four inches of the tube) and update the bases for this criteria. This proposed change willnot affect repair criteria for Baw indications located outside of the tollexpansion region. The proposed Hcense amendment willalso modify the current requirement that 'The plugging limitfor sleeves willbe determined prior ta the 1987 refueling outage for Cycle 15," to,
'The plugging limitfor sleeves willbe determined prior to the Brat refueling outage followingsleeve installation,"
since. to date, no sleeves have been installed at the Haddam Neck Plant. The proposed license amendment willalso revise Technical Specification Section 4.10.D~ to delete the exclusion of tube row 37. column 73 in Steam Generator 2 froin plugging. since this tube has subsequently been plugged during a mid-cycle shutdown in July, 198L Basis forproposed na significant hazards consideration determfnationt In accordance with 10 CFR 50.92, the licensee has reviewed the attached proposed changes and has concluded that they do not involve a significant hazards consideration. The basis for.'this conclusion is that the three criteria of10 CFR 50,92(c) are not compromised; a conclusion which fs supported by the licensee's determinations made pursuant to 10 CFR 50M, The proposed changes do not involve a signfficant hazards consideration because these changes would not:.,
- 1. Involve a significant increase in the probability or consequences of an accident previously analyzed. As stated above, since these changes meet the safety margins ofRegulatory Guide 1.121, the probability of the design basis accidents remains unchanged. Sfnce the leakage is less than allowed by the existing technical specffication. the consequences of the design basis accidents are not impacted.
- 2. Create the possfbilfty of a new or different kind of accident from any previously. evaluated. Na new systems or designs are introduced by these proposed changes; therefore..na new failure modes are createcL In adcH tiaa.
the plant operating characteristics remain unchanged. Thus, aa new accidents are created.
- 3. Involve a significant reduction fn R margin ofsafety. As stated above, the proposed changes wauld maintain existing structural margins and the total leakage (primary to secondary) would
'emain within the acceptable Hmita of the existing technical specfficatfan.
Thus, these changes wiffnot reduce the margin ofsafety as defined in any plant technical specification.
The staff has reviewed thc licensee's detennfnatfcm that the proposed Hcease amendment involves no sfgnfficaat
'axarda considerations and agrees with the licensee's analyses. Accordfn@r, the staff proposes to determine that the
'roposed license amendment does not involve significant hazards considerations.
Local Public Document Roam location: Russell Library, 123 Broad Street, Middletown, Connecticut 08457.
Attorneyforlicensee: Gerald GarfiefcL Esquire, Day, Berry and Howard, Counselors at Law, City Place, Hartford, Connecticut 081034499.
NRCProj ect Director. Cecil O.
Thonlas Consumers Power Company, Docket No.60-255, PaHsades Phmt, Vaa Buren County, Michigan Data ofamendment request: June 5,
" 1987 Description ofamendment request:
This request modifies a previous request for a change in the augmented fn-senrfce inspection requirements for the steam generators dated September 28, 1984 (50 FR 20975). This request would maintain the interval for such inspections at each refueHng but not to exceed 24 months except that the Intenral could be extended to 30 months provided the mean degradation increase durfng the previous inspection interval was less
~
than 1 percent I
Basis forproposed no significant hazards consideration detarminatiom Inspection results, that show for the previous operating interval that there was essentially no degradation of steam generator tubes, would ensure that the probability and consequences of the accident previously evaluated in the FSAR, Le steam generator tube rapture, are unchanged. The possibility of an accident not previously analyzed fa nat created because this change only affects steam generator tube integrity, and the rupture of these tubes has beea previously analyxccL Since the inspection results sheared no sfgafBcaat degradatfan In Qsc preccfoua fntenraL ao sigaffjcaat redaction ie thc aMrtfha.af safety ia caused by this exteasiaaL Based on the farcipimg. the staK propoaea to determfne that'he proposed change invalves no significant hazards.
consideration.
LocalPublic Document Boom
, lacation: Van Zoeren Library, Hope College, HaffancL Michigan 49423.
Attorneyforlicensee: Judd I Bacon, Esquire, Consumers Power Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Project Dr'rector. Martin J.
Virgilio,Acting.
Dalryland Power Cooperative, Docket No. 50409, LaCrosse BoilingWater Reactor, LaCrasse, VIGsconsfn Bate ofamcrxfmant request: May 22, 1987 Descriptio ofamendment reqaeaa The Hcensee proposes that License No.
DPR45 for the LaCrosse Bcnling Water Reactor (LACBWR)be amended to possess-bnt-not-operate status. The licensee stated In a letter dated AprZ29, 1987 that LACBWRwould be permanently shut down and a decommissioning plan submitted to the NRC. LACBWRwas permanently shut down on April30, 1987.
'aaia forproposed no significant hazarda consideration detanninatian..
The Commission has provided gufchince concerning the appHcatfon of the standards for determining whether a.
significant hazards coasideratfon exists in 10 CFR Part 5082 by provfding certain examples (51 FR 7751), One of the examples (ii)of actions not likelyto involve a aignificant hazards consideration relates to changes that constitute additional restrictions or controls not presently included in the license.
The proposed action to amend Lfcensct No. DPR45 to possess-but-not-operate status is more restrfctive than the present license because the present license would permit operation. ofthe facility.Therefore, since the proposed amendment is encompassed by example (ii)ofactions that are consfdered not likely to involve aignificant hazards conskferatfon, the Commfsshn hcs made a proposed determination that thc proposed action does not involve a significant hazards consideratioa.
Local Public Document Racpm location: LaCroua Public Library, 800 Main Street, LaCroase, Wfsconsin 54801.
Attorneyforlicensee: Kevin Gallen, Esquire, Newman and Holtzfnger. 1015 I, Street, NW, Washington, DC'20038 NRC Prbject Director. Herbert N.
Berkow'
I Fedecd RegbAer / Vol. 52, No. AS I Nedaeaday, July 1, 1887, f'Notfces Detroit Edfson Coaspany, Docket No. 58.
341, Fermi-2, Monroe County, MkLlym Date ofamendnient request: January 28, 1987, as superseded May 28, 1987
" Description ofamendment request:
The'proposed an'tenthnent would revise the Fermi-2 FacilityOperating License ¹. NPF-
- 43. License Condition 2.C.(10) eiitftied, "Emergency Diesel Generator Lube Oil Surveillance Program (Section 9.5.7, SSER 5)", to: (1) incorporate the requirement forpeifodfc gap checks of the Emerymcy Diesel Genirator (EDG) engine main bearings; and (2) delete the requirement for the disassembly and removal of oil filters and substitute the requirement for a monthly analysis of EDG engine lube oil samples. The inspections and analyses required in the revised license condition will supplement the action and surveillance requirements pertaining to the EDGs fn Section 3/4.8.1 of the Fermi-2 Technkal Specifications (Appendix A to NFL).
The license amendment proposed by the licensee's May 28, 1987 letter supersedes the licensee's earlier proposal dated January 28, 1987, and was submitted in response to the Commission's safety evaluation dated April7, 1987.
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) inv'olve a significant increase in the probability or conseqaences of an acddent 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 Commission's staff has reviewed the licensee's January 28, 1987 application as supplemented.
and has determined that the proposed amendment involves no significant hazards consideration for the reasons stated below:
(1) The initialEDG engine lube oif surveillance program referenced in the present license condition was developed as a result of bearing failures which occurred in January 1985, and required the quarterly disassembly and inspection of the engine labe ofi fifter.
Evaluation of data obtained by the licensee in the perfonnance of fdter disassembly and Inspec6on (submitted by the licensee's January 28. 19K letter) fndlcates that thfs fjispectfost method fs ineffective and fncondasfve ia predictfng fndpfent fsffsre. Aa augmented bearfn6 surve ce prognnn fnstitated by the licensee as a resaft of subsequent bearing failures in late 198$,
and not encompassed fn the present ',:
license condition, is considered to be a more direct and effective means foe detecting indpient bearing failure, This method involves the periodic measurement of the main bearing gapa in each engine. The proposed revised license condition incorporates the bearing gap check inspection. The propose'd license condition also continues to require periodic analysis of lube oil filtersamples (without filter disassembly and removal), which the staff agrees willreduce the probability offilterdamage and the'Introduction of foreign material into the lube oil system which could eventually damage engine bearings. The revised license condition willrequire periodic inspectfo'ns which willmore effectively provide indications of bearing performance and "
,consequently better ensure the availability of the EDGs when needed.
Therefore, the proposed license condition amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
(2) As stated in (1) above, operating experience has shown that the frequent disassembly of the oil filteras currently required has not been effective in predicting incipient bearing failure.
Further. the frequent disassembly of the oil filterincreases the likelihood of damaging the filterand/or introdudng foreign material Into the lube oil system that could possibly damage the engine bearings. The addition of the bearing gap checks as a part of the revised proposed license condition is considered to be a more positive method of determining bearing. performance, minimizing the occurrence ofbearing failure. and better ensuring the availability of the EDGs when needetL Therefore. the proposed license condition amendment does not create the possibility ofa new or different kind of accident from any previously evaluated.
(3) As stated in (1) and (2) above, the new license condition, which incorporates bearing gap checks as an inspection requirement. and retains the requirement for the analysis of engine lube oil samples, willincrease the margin of safety, providing more reliable information on bearing perfonnanoe, wear, etc. As such, the proposed license condition amendment does not involve a signfifiicant redactfoa on the margin of safety.
, The COQunfssfon proposes to
'determine that the revfsed license condition proposed by the licensee does not involve a significan hazards cons Meration.
~ Local Public Document Room location: Monroe County Library System, 3700 Custer Road, Monroe, Michigan 48161.
~ Attonieyforthe licensee: John Flyn, Esq., Detroit Edison Company, 2000 Second Avenue, Detroit, Michigan 48226.
NRCProject Directon Martin J.
VirgiHo,Acting.
Detroit Edison Company, Docket No. 50-341, Fermi-? Monroe County, Michigan Date ofamendment requestt May 27, 1987 (NRC4740N)
Descn'ption ofomendment request:
The proposed license amendment would change the Fermi-2 Facility Operating License No. NPFM, Technical Specification 3/4.8.2, Table 4.8.2.1-1 entitled, "Battery Surveillance Requirements" to delete Table,.
Notations (7) and (8) and the applicable Bases which specify battery surveillance parameters for a nominal spedfic gravity electrolyte of1~. These parameters are no longer considered necessary by the licensee due to the replacement of the Division H batteries with cells containing 1.210 nominal specific gravity electrolyte.
Basis forproposed no eignificant hazards c'onsideration determination:
The Commissiori 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 acc'ordance with the proposed amendment would not: (1) involve a sfgnificant increase in the probability or consequences of an acddent 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 determined that the
, changes proposed to Technical Spedficatfon 3/4.8.2, Table 4.8.? 1-1:
(1) Would not involve a slgnfficant increase hi the probability or consequences of an accident prevfously evatuated since the parameters. for a norainaf spedfic gravity electrolyte of 1.250 have been superseded by the parameters for a nominal specific gravity electrolyte of1.210, which are already spedfied in Technical Specification Table 4AL2.1-1 of the Plant Technfcaf Spedficatfons. 'Hie change is an editorial correction and
Federal Register / Vol. 52, No. 126 / Wednesday, July 1, 1987 / Notices administrative in nature and therefore falls into the category of amendments that are considered not likely to involve significant hazards consideration (51 FR 7751). Deletion ofTable Notations (7) and (8) does not involve a physical change to the facility, change a limiting condition of operation (LCO), or change any operating practice; nor does removal ofTable Notations (7) and (8) change any safety analysis or design basis at Fermi-2.
(2) Would not create the possibility of a new or different kind ofaccident from any accident previously evaluated. As stated in (1) above, the prop'osed change is administrative in nature and the r moval ofTable Notations (7) and (8) from Table 4.8.2.1-1 of the Technical Specifications does not involve a physical change to the facility, change an LCO, change any operatin'g practice, or change any safety analysis or design basis at Fermi-2.
(3) Would not involve a significant reduction in a margin of safety. As stated in (1) and (2) above, the proposed change is administrative,.and the removal ofTable Notations (7) and (8) from Table 4.8.2.1-1 of the Technical
'pecifiications does not involve a physical change to the plant or its operation. or any safety analysis or design basis which would cause a s'gnificant reduction in safety.
The Commission has reviewed the licensee's no significant hazards consideration determination and agrees with the licensee's analysis, Accordingly, the Commission proposes to determine that the requested.
amendment involves no significant hazards consideration.
Local Public Document Room location: Monroe County Library System, 3700 South Custer Road, Monroe, Michigan 48161.
Attorneyfor the licensee: John Flynn, Esq, Detroit Edison Company, 2000 Second Avenue, Detroit. Michigan 48228.
NRC Project Directorr Martin J.
Virgilio,Acting.
Duke Power Company, Docket Nos. 50-413 and 50-114, Catawba Nuclear Station. Units 1 and 2, York County, South Carolina Date ofamendment request: June 10, 1987 as supplemented June 11 and 16, 1987 Description ofamendment requestt The proposed amendme'nts would allow the extension. on a one-time basis, of several 18-month Technical Speciifiication (TS) surveillance intervals until the first refueling outage for Catawba Unit 2. This extension is needed because these surveillances can only be performed with the Unit in Hot Shutdown (Mode 4), Cold Shutdown (Mode 5), or Refueling (Mode 6).
Although the proposed amendments were requested for both Units 1 and 2.
changes are proposed forUnit 2 only.
Unit 1 is included in this notice only because the TSs are combined in one document for both Units.
Normally, since refueling outages occur about every 18 months. extension beyond the 18-month interval required by the TSs for such surveillances is usually not necessary. However, due to the extended length of the Unit 2 startup program and cycle 1, the licensee must either req'uest and receive an extension or shut down prior to the first scheduled refueling outage. Similar extension was approved for Catawba Unit 1 by amendments issued July 3, 1986 (Amendment No, 8 for Unit 1 and No. 1 for Unit 2). Unit 2 is currently scheduled to enter its first'refueling outage on December 30, 1987. Most of these surveillances must be performed on August 15, 1987, or later. Therefore, the longest extension entails a period of4.5 months, Furthermore, the tests required willbe performed ifan outage of sufficient duration occurs prior to the first scheduled refueling outage.
The particular surveillances and the time at which the surveillance interval (including the 25% grace period allowed by TS 4.0.2) willexpire are discussed
- below,
~
1, Feedwater Isolation on receipt ofa high doghou'se water level signaL TS Table 4.3-2, item S.d. The trip actuating device operational test would be extended from August 15, 1987, and would be performed prior to entering startup (Mode 2) or Hot Standby (Mode 3), as applicable, followingUnit 2 first refueling. There have been no failures of this circuitry and no actuations since preoperational testing.
- 2. Turbine Trip on loss of all main feedwater pumps, TS Table 4.3-2, item 6.d, the trip actuating device operational test'would be extended from August 15, 1987. and would be performed prior to entering Startup (Mode 2) or Hot Standby (Mode 3), as applicable, followingUnit 2 first refueling outage.
This instrumentation is reliable and has operated satisfactorily due to one challenge after completion of preoperational testing.
- 3. Turbine Trip on reactor trip, TS 4.3-2, item B.e. The trip actuating device operational test would be extended from August 15, 1987, and would be performed prior to entering Startup (Mode 2) or Hot Standby (Mode 3), as applicable, followingUnit 2 first refueling. This instrumentation ls reliable and has responded satisfactorily in response to seven challenges.
- 4. Turbine Trip on steam generator water level-high-high, TS 4.3.2.2, Table 3.3-5, item 7.a. The response time test tytiuld be extended from August 15, 1987, and would be performed prior to entering Hot Shutdown (Mode 4) followingUnit 2 first refueling. This instrumentation is reliable and has responded satisfactorily in respone to three challenges.
- 5. Feedwater Isolation on steam generator water level-high-high, TS 4.32 2, Table 3.3-5, item 7.b. The response time test would be extended I'rom August 15, 1987, and would be performed prior to entering Hot Shutdown (Mode 4) followingUnit 2 first refueling. This instrumentation is reliable and has responded satisfactorily in response to three challenges.
- 6. Feedwater Isolation on a reactor trip coincident with low reactor coolant system average temperature, TS 4,3.2.2, Table 3.3-5, item 8. The response time test would be extended from August 15, 1987, and would be performed prior to entering Hot Shutdown (Mode 4) followingUnit 2 first refueling. This
'nstrumentation is reliable and has responded satisfactorily in response to three challenges.
- 7. Turbine~ven AuxiliaryFeedwa ter Pump Steam Supply Valves Surveillance to verify that they open upon receipt of an auxiliary feedwater actuation test signal, TS 4.7.1.2.1b.3). The test would be extended from August 15. 1987, and would be performed prior to entering Hot Standby (Mode 3) followingUnit 2
.first refueling. These equipment and instrumentations are highly reliable and have responded successfully in response
- to ele'ven challenges,
- 8. Containment Valve Injection Water System Surveillance to verify injection flow to containment isolation valves, TS 4.6.8.2. The test would be extended from, October 29, 1987, and would be, performed prior to entering Hot Shutdown (Mode 4) followingUnit 2 first refueling. This system is reliable with a
'ood operating history and only three valves have not been tested.
- 9. Diesel Generator Inspection, during shutdown, in accordance with the manufacturer's recommendations, TS 4.8.1.1.2g.1). The inspection would be extended from January 9. 1988. and would be performed prior to'entering Hot Shutdown (Mode 4) followingUnit 2 first refueling. A complete inspection in accordance with the TDI diesel generator owners group inspection prograin was conducted on Unit 2 engines with satisfactory results.
Fradazal Reghler / VoL 52, No. 120 / Nedrtesday, Jaly 1, tMF / N08ces Routine sarveilhmce activttfes wgt continue to be coadected on schehde.
- 10. System Response Time tests for the primary RTDs assodated with the Overtemperature Delta Tand, Overpower'elta T Reactor Taps, 'IS 4.3.1.2. Table X3-2. items 7. aad LThese tests woatd be extended from September 11. 1987, aad would be performed prior to entering Startap (Mode 2) followingUatt 2 ffrst refaeliag.
Only the RTDs remain to be tested. 'Ibe rest of the ctrcuthy has been tested satisfactorily wt0dn the requhed surveillance interval. These RTDs are reliable and have successfully,met the required response time during the last three previous tests (two oa Unit 1 and one on Unit 2).
Basis forpIaposed no significant hazards cansiderat&n determinatian The Commission has provided certatn examples (51 FR 7744) of actions Bkely to involve no signiffcant hazards considerations. The request involved ta this case does aot match aay of those exampleL However, the staff has reviewed the hceasee's repsest for the above amendments and determined that should this request be implemented, tt would not (1) involve a stgatffcant increase in the probaMity or consequences ofaa acddent previously evaluated. The probaMity of an accident is not sigaiffcantly increased because these ctumges willaot affect the design or operation of the Uait. The consequences ofan acctctent willnot be signiffcantly tacreesed since the systems affected are required to be operable through other applicable TS requirementL Furthermcire, the extension to the survetttance intervals ts for a brief period (4.5 months), and should not stgniffcantty affect the ebiBty of the sytems to functian properly. Also, it would not (2) create the possibility of a new or dttIereat kind ofacddeat from any acddent previously evaluated because the design and operation of the Unit wttt aot be affected. Therefore, no new kinds of acddents are introduced.
Finally, it would not (3) involve a
~
significant reduction in a margin of safety because the surveillance interval extension is for a brief period (LS months), and the systems are required to be operable through other applicable TS requirements. In addition, the equtpmeat has proven to be reBable thraugh satisfactary'responses to actuations and
~ prior inspections.
,,Based an the above, the Commission proposes to determine that the changes do not involve stgniTicant hazards
'onsiderations.
Local Public Dacurrrrant Bacrn Axation: York County Ltbczay, 13$ East Black Street, Rock HN, South Catalina 29730 Attorneyforlicensee: Mr. Albert Carr, Duke Power Company, 422 South Charch Street, Chadotte, NorSL CaroBna 28242 NBC Prelect Duectar: Ej. Ycamgtsbosl Daquesne Qgbt Company, Dctcket No.
6MS4, Beaver Valley Power Station, UnitNa 1, Shtpptagpart, Pennsylvania Dote ofcuaendarent rertuest: AprQ 13.
1987 Descrr'pliua ofcunendceent neqrrest:
The proposed ameadmeat would remove technical spectftcattoas of the Waste Gas Decay Teak Monitor, aad revise certain surveillance requirements
, as fol1ows:
- 1. Table 33-13, ddete items 43) aad 4.c which reference the moaitar and the associated sampler Qow rate measuring devtceo
- 2. Table 3&13. delete Action 35 since this action statement only appQas to the monitor.
- 3. Table 43-13. delete items 4h aad 4.c to reQect the change to Table 3W13.
- 4. SurveQlance requiremeat 4.112.L1, would be darified by modifying reference to the Waste Gas Decay Tank
- Monitor,
'asis forproposed na significant hazanis cansidetatian determination:
The Comadsston has provided standards for determining whether a stgntffcant hazards consideration exists (10 CFR 80.92(c)). Aproposed amendmeat to an operating Bcense fora facilityinvolves no slgntffcant hazaats constderattan tfaperation ofthe fadBty in accordance with the proposed amendment would not (1) hnrolve'a stgatffcant hicrease tn the probability or consequences of an acddent previously evaluated; (2) create the posstMtty ofa new or different kind af acddent from any acddent previously evaluated; or (3) tavobre a significant redaction in a margin ofsafety.
The Bcensee stated that the proposed changes do not involve a stydttcant hazards nmstderatton because:
- 1. The revision to surveillance requhement 4.11.2.5.1 would requhe verification of the quantity of radioactive material contained in each gas storage tank at least ance per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when radioactive materiats are being added to the tank and,the concentration ofthe primary coolant ts greater than 100 micro carie/ml. The licensee stated that this action would ensure the acddent analysts value in the FSAR would not be exceeded aad that NUIKGO472. Revision 2 does sot Bat a radiation aec6ior or a mnpter Seer rate moasariag ctevtca oa the waste gaa hotdojr syatesa.'Ibe ttoeasee stated thert operation ofthe monitor provides no additional ted'annattan that waaM increase the level'of safety, Therefore, deleting this monitor fram the techdcal spectffcattans wouM not affect the prab*Btty'afocaareace ar the consequences ofan acchhmt prevlausly evaluated.
2.'Ihe proposed changes are bounded by the FSAR waste gases release
~
acddent analysts (Section 14~),
SpectffcaQy, proposed surveillance requirement 421~ would easure that appropriate sampBag is performed so that the FSAR acddeat analysis results would not be reached or exceeded.
'I%us, na adverse safety constderatioas would be introduced by this proposed change to the technical spedffcattons.
Therefore, the probability for ea acddent or malfunction ofa type different fmm the previously everted waste gas release acddent would not be created.
- 3. SpectBcatton 3.1~ would restrict the quanQty ofradioactivity coataiaed ia each gas storage tank. This would provide assurance,that ia the event of an uncoatroQed release of the teak's contents. the resulting total body exposure to aa iadtvlchuLl located at the nearest exchsiaa area boundary for two hours taaaedtately fallowingthe oaset of the release would aot exceed L5 rem.
The spedfiad Bcatt restricting the quaattty ofradioactivity coatained ia each gas storage tank was specified to ensure that the total body exposare resulting frcaa the peotshted release remained a scdtable fraction of tha refereace value set forth inICFR 100,11(a)(Q. Theasfora. the proposed changes would not af5ect the awugtn of safety aad woutd be cosstetent with the FSAR acddent analyses.
We staÃcancms with the licensee's assessment and proposes to determine that the requested amendment involves ao stgatRcaat hnzants consideratioa.
Local PirMcDacrrrnent Baanrr locatiara S.F. Jones Memoria'Ltbraty, 803 PraakBa Avenue, ABqutppa, Pennsylvania lSXH AttorneyforIkenseer Gerald Charnow. Esquire, Jay E Sttberg, Esquice, Shaw, Ptttman. Potts, and Trowbridge, 2300 N Street, NV, Washington. DC 2003tr NitCPraj eat Director: Jolm F. Stot Daquisae Light Ccaayaay, Docket Na.
6M34, Beaver Vasay Pewer Stattaa Uatt No. 1, Shtpptagpcat, Peaasyivcatct Date ofamendment request April29, 1%7 Briefdescnpherr ofamend&ant request"Ihe Technical SpecHicattans for'Beaver ValleyUnit 2 were developed
- 24660, Federal,Register / Vol. 52, No. 128,/ Wednesday, July 1, 1987 / Notices based on those for Unit 1. In the process of development, the staff and the, licensee discovered errors and needs'for clarification in the Unit 1 Technical Specifications. Errors that involve technical review are being addressed by the licensee and the itaffseparate from the subject request.
The requested amendment would correct editorial errors (spelling, capitalizati'on, grammatical etc.) and restate some specifications in the same way they are stated in the Unit 2 Technical Specifications. In addition, the amendment would relocate license, condition 2.C(6), concerning secondary water chemistry monitoring program, to Section 6.8.5 of the Technical Specifications. This relocation does not change the nature of the requirement and is only an editorial charige.
Basis forproposed no significant hozards consideration determination:
~ The Commission has provided guidance concerning the application of these standards by providing certain examples (51 FR 7751). One of these, Example (I), involving no significant hazards considerations is "Apurely administrative change to technical specifications." The requested changes all match this example. On such basis, the staff proposes to characterize these changes as involvingno significan hazards consideration.
Local Public Document Boom location: B.F. Jones Memorial Library, 663 Franklin Avenue, Allquippa, Pennsylvania 15001 Attorneyforlicensee: Gerald "
Charnoff. Esquire, Jay E Sifberg, Esquire. Shaw, Pfttman, Potts, and Trowbridge, 2300 N Street, NW, Washington. DC 20038 NBCProject Directon John F. Stolz GulfStates Utilities Company, Docket No. SM58, River Bend Station, Unit 1 West Felfcfana Parish, Louisiana Date ofamendment request: March 10, 1987 as supplemented June 9, 1987.
Description ofamendment request:
The proposed amendment would modify the Technical Specifications (TSs) to extend the surveillance interval for certain Surveillance Tests. The proposed amendment would revise the Technical
"'pe'cifiications as follows:
(1) TS 4.4.2.1.1(b) would be changed to require a channel calibr'ation of the acoustic monitor for eac~ safety/relief valve to be performed at least once pe'r refueling cycle instead of at least once per 18 months as required in the current TS so (2) TS 4.6.2.2 would be modified, on a one-time basis, to extend the surveillance schedule for testing the drywell bypass leakage by three (3) days until the first refueling outage scheduled to begh on September 15, 1987.
(3) TSs 4,4,3,2,2a, 4,6,1.34 4A41.3f, 4.6.1.3i, 4,6.2.3.d2and 4.6.2,1.ed would be modified, on one-time basis, to exten'd the schedule for the required surveillance leak rate tests by a period of time that'ranges from 5 days to a maximum of 41 days until the fIrst refueling outage scheduled to begin on September 15. 1987, The current Technical Spedficatfons require a surveillance test to be performed at intervals no greater than 24 months for TS 4.6.1,3d "and at intervals no greater" than 18 months for the other TSs that are proposed to be modified. The types of leak rate tests involved are Type B and C tests spedfied in Appendix J to 10 CFR Part 50, reactor system boundary valve tests, and tests of air systems providing a seal against diywelf bypass leakage. The request for extensions of the surveillance interval fncludes a total of 52 valves. An exemption to Appendix J to 10 CFR Pait 50, Section BLD.3 is also required. This section requires Type C testing ofisolation valves at intervals not to exceed 2 years, which is the basis ofSection 4.6.1.3d'of the TSs.
The licensee's request for a one-time extension of the surveillance schedule for the primary containment/drywefl hydrogen mixing trains fs the subject of a separate Federal Register notice.
Basis forproposed no significant hazards consideration determination:
The Commission has provided standards for determinhg whether a signfficant hazards consideration exists as stated in 10 CFR 50.92(c). Aproposed 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 ofan accident previously evaluated; or (2) create the possibility of a new or differ'ent 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.
(1) Change in the surveillance interval for performing a channel calibratfon of the acoustic monitor for each safety/
relief valve. With regard to the three standards, the lfcensee stated;
- a. The proposed change does not involve a significan increase Iiithe probability or conaequencea ofan acddent previoualy evaluated because the change in the surveillance Inteivaf willnot result In a decreaae In the instrument accuracy. Thfa poaition Ia supported by the msnufactureA information that there ere no time refatei drifteffects on the fnatrumant. as s result the extension willnot result in a decrease In the Safety Relief Velve acoustic monitor accuracy and therefore, the'response to
'previously evaluated events wfilbe iiiichanged. Since the revfaion does not hvofve a design, configure tion or i operational change to the plant and the response to events Ia unchange*There Ia no Increase in the probability or consequence of any acddant previously evaluated.
- b. The proposed change does not create the possibility of a new or dNerent kind of scddent from any accident previously evaluated becauae the change in the frequency of calibration of the acoustic monitors ia coriafatent with tlm deaign apedficatfon for the system and the safety analysis, and does not Involve a'design change or physical change, aiid therefore does not alter the dealgn response of the instrumentation. Thus, no new acddent scenario ia introduced by this revised frequency ofcalibration of the acoustic monitors.
- c. The proposed change to the scouaiic monitors surveillance period does not Involve e significant reduction In a margin of safety because the change in the frequency of calibration for the acoustical,monitor aenaor Ia conaiatent with the design apedfication of the system and the sensor design Ia not sensitive to the surveillance period.
Additionally, as delineated in the juatificatfon, the proposed change willnot effect the performance requirements In the LimitingConditions of Operation contained in the Technical Spedficatlona. Thus, the margh ofsafety Ia not impacted.
(2) One-time modification of the surveillance interval for testing the drywell bypass leakage. With regard to the three standards. the licensee stated:
- e. The proposed amendment to the Technical Spedficationa would not involve a aiydficant Increaac In the probability or consequences of an ecddent previously evaluated because there Ia no change In the dealga or performance ofplant ayatema or components from those evaluated fn the Final Safety Analysis Report (FSAR). The proposed revision Ia consistent with the acddent analyaea described In the FSAK Due to the near passive nature of the drywall structure, allowing the DrywallBypaaa Leak Test to be performed at the refueling outage willnot result in any additional foaa of structural Integrity. Test data previously obtained reveals that the current drywell bypass leakage Ia a amaff fraction of the allowed leakage hence removing further the
'ossibility of exceeding the design anelyaia; No incresaa In the probability or conaequencea'of an ecddent, therefore,.
exfata
- b. The proposed change does not create ihe possibility of a new or different type of acddent from any acddent prevfoualy evaluated because this change does not Involve a dealgn change or Involve a change In the operatiiig mode ofexisting equipment.
Thus. no new accfdant scenario Ia Introduce*
- c. The proposed change does not involve a significant reduction In the margin of safety becauae the margin of aefety diacuaaed in the
Federal Register / Vol. 52, No. 128 / Wednesday, July1, 1987./ N08ces '4551-FSAR Section L2,.1.1.3 4 assumes the drywell bypass test to be conducted at each refueling outage. Allowingthe technical spectficatlon to ayee with the statement in the FSAR does not, therefore, cause a reduction in the margin of safety previously evahiated. Item a above furthermore, discussed the passive natuie of the drywell and itis because of this passive nature that it can also be stated that no reduction In the margin ofsafety exist.
(3) Change in the surveillance intervals, on a one-time basis, for performing leak rate tests. With regard to the three standards, the licensee stated:
- 1. The proposed amendment would not involve a significant increase in the probability or the consequences of an accident previously evaluated results from this change because:
- a. The valves were last tested satisfactorily and due to the short period of the extension.
no significant increase in the probability of
'quipment failure ls postulated.
- b. The LLRTtesting provides verification of valve seating Integrity and does not provide assurance of the actuation of the valve when called on to perform its isolation function.
The increase in surveillance frequency does not affect the probabilities of the valve actuating when called upon to perform its required isolation. Isolation function testing are satisfactory and cunent.
- c. The change Increases allowable surveillance interval less than 8% beyond the current conservative surveillance requirements and has no affect on the assumptions ofvalve leakages assumed In the present accident analysis.
- z. This change wou!d not create the possibility of a new or different kind of accident from any accident previously evaluated because the proposed change introduces no new systems, modes of operation, failure inodes or other changes to any equipment. The proposed change'oes not change the system functional analysis and therefore, new accident scenarios are not credible based on scheduling of testing alone.
- 3. This change woidd not involve a significant reduction in the margin ofsafety because, based on the enclosed technical justification which indicates the number of valves and penetrations Involved. their current leakage rates and estimated leakage rates at proposed 09-1547 refueling date, the.
following can birstated:
- a. Those valves for which extensions are being requested have for the most part. based on initial and subsequent LLRTresults, exhibited a high degree of leak tight reliability.
- b. Overall LLRTshows a very tight containment.
- c. The drywell airlock and the personnel door in the drywell equipment hatch have
'demonstrated a high deyee of leak tight reliabilityand are infrequently used.
- d. The requested extension does not significantly increase the allowable frequency interval provided in the Technical Speciiications. (The maximum increase is approximately,8%.)
- e. There willbe no Identified increase in postulated Individual offsite or cumulative occupational radiation exposure as a result of the requested amendment which merely requests to delay testing.
- f. The requested amendment concerns schedule relief for surveillance testing ofa limited number of contahment isolation valves and drywall access doors willnot result in a significant change in the amounts or types ofeifluents that may be released off-.
site.
I The NRC staff has reviewed the licensee's no signlflcant hazards consideration determination and agrees'ith the analysis, Local Public Document Boom Location: Government Documents Department. Louisiana State University.
Baton Rouge, Louisiana 70803 Attorneyforlicensee: Troy B. Conner,.
Jr., Esq., Conner and Wetterhahn. 1747 Pennsylvania Avenue, NW, Washington, DC 20008 NBCPleat Director: Jose A. Calvo GulfStates UtflitlesCompany, Docket No. SM58, River Bend Station, Unit 1 West Felldana Parish, Louisiana Date ofamendment request: March 18, 1987 as supplemented June 9, 1987.
Description ofamendment request:
The proposed amendment would revise the Technical SpecHications (TSs) to extend the surveillance intervals for the automatic depressurization system (ADS) and the calibration frequency of the drywell air cooler condensate flow.
The proposed amendment would modify the TSs as follows:
(1) TS 4.5.1.e.1, 4.3.3.1-1 &2.h, 4.3.3.1-1.A.2.i, and 4.3.3.2 would be changed to extend the surveillance interval for the ADS, on a one-time basis, to the refueling outage scheduled for September 15, 1987. The surveillances willbecome overdue on August 18 and
- 18. 1987 in accordance with the current TSs which require that the surveillances be performed once per 18 months. The TSs require manual Initiation and functional testing of the ADS.
(2) TS 4.4.3.1,c wou14 be changed to extend the surveillance interval for
performing the drywall air cooler condensate flowrate monitoring system channel calibration from at least once per 18 months to at least once per 24
- months, Basis forpraposed na significant hazards consideratian determination:
The Commission has provided standards for determining whether a siydficant hazards consideration exists as stated in 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 difKerent kind of accident'from.
any accident previously evaluated; or (3)
~
involve a slgnlflcant reduction in a'a~
ofsafety. The licensee addressed the above three standards in the amendment application.
(1) One-time extension of the surveillance interval for the ADS. With regard to the three standards, the licensee states:
- a. The proposed change does not Involve a siydficant increase in the probability or consequences ofan aciddent previously evaluated because the increase in the surveillance interval willnot result in a reduction in system reliabilitynor willit effect the abilityof the system to perform its design function. This change wiflnot effect system configuration or operation. This change in surveillance interval is supported by successful completion ofstart-up test.
successful performance of SRVs during plant, operation. the ldgh reliabilityof the system components, and by completion of monthly functional testa ofsystem components as required by Technical Speidficattons.
- b. This change willnot create the possibility of a new or different ldnd of accident from any accident previously evaluated because it does not Involve any changes to system conliguration or operation.
Achange in surveillance interval willnot create any new a'cclderits.
- c. The proposed change wIIInot siydficantly reduce a margh of safety because the reliabilityofthe system to perform its function Is not significantly effected. The system design. operation. and ability to function when reqidred remain unchangecL Additionally,as delineated in the justlficatlon, the proposed change willnot effect the performance requirements in the LimitingConditions of Operation contained in the Technical Specification. Thus, the margin ofsafety Is not impactecL (2) Extension of the surveillance interval for performing the drywell air cooler condensate flowrate monitoring--
system calibration. Withregard to the three standards, the licensee states:
- a. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because the change in the surveillance interval willnot result in a decrease in the required Instrunient accuracy.
This position Is supported by the purpose of this,system is to alert and aid the operator In defining an event, no credit Is taken In the Safety Analysis for the alarms for high drywell air cooler drain flow.Operator action results from information received from the drywall sump flownot the air cooler condensate flow and therefore, the response to previously evaluated events willbe unchanged. Since the revision does not involve a design, configuration or operational change to the plant and the response to events is unchanged. There ls no increase in the probability or consequence of any accident previously evaluated.
- b. The proposed change does not create the possibility ofa new or cfiiferent kind of accident from any accident previously
Fodoral Reghler f Vol. 62, Na. 126 / Wednesday, Qy 1, f987 I Notfcaa evaluated because the change in the frequency of caIIbratha of the drywall ait cooler drain fiow Indication wIIItata'
'ithin the present design criter and ',;,.;
respond as previously evaluated and docs
not Involve a design change or physical change, and therefore does not alter the deafgn response of the Iriattumcntatimt.'Isa, no new accident acermrio ia Inlroduecd by this revised frequency afcallbrstha ofthe acoustic monitors.
c.,Thc proposed change to the sutv'citiaacc period docs not Involve a aignificant reduction ln a margin of safety because the change in the frequency of calihtattaa for the drywall air cooler drain fiowiadicatiim ia not used'in the accident analysis and thctcfoce no tcducticm Ia the analyacd margin of safety willbe crsatccL Additionally. aa dcHncated Ia the Iuatificatlon,'he proposed change willnot effect Ibc performance requirements In the
'Imittog Conditions of Opctathn amtninad ia Ihc Tcchnical Spceificatioa. Thus, the marsin of safety Ia not impactccL The'staff has reviewed the licensee's no significant haxanls consideratioa determination and agrees with the analysis.
Local Public Document Room Location: Government Documents Department. Louisiana State University, Baton Rouge, Louisiana 70803 Attorneyforlicensee: Troy B. Conner, Jr., Esq., Conner and Wetterhahn, 1747 Pennsylvania Avenue, NW, Washmgtoa, DC 20008 NRC Prej ect Director: Jose A. Calvo Gulf.States Utilities Coctipaay, Docket No. 50458, River Bend Statha, Unit 1 West Fcliciaaa Parish, Lacdsfaaa Date ofamendment request: May 11.
1987.
Description ofanrendment r'equeet:
The proposed amendment would revise the technical specifications (TSs) to extend the surveillance Intervals, oa a one-time basis, for the penetration valve leakage control system. the main steam line radiation-high isolation actuation instrumentation, and DivisionI and Division II18 month ECCS surveillance tests. The proposed amendment would modify the TSs as follows:
(1) TS 4.6.1.10.c would be modified ta extend the interval for performing the functional test of the penetration valve leakage control system, on a one-time basis, from every 18 months until the first cycle refueling outage scheduled to begin September 15, 1987. The current surveillance overdue date is September 8, 1987: hence, the extension is for a period of up to 7'days.
(2) TS 4.3.1.1, 4.3.'1.2, 4.3.Z.I, 4.3.2.2, and Tables 4.3.1.1-1 and 4.3.2.1-1 would be modified to extend the interval for performing the main steam line radiation-high channel calibration (Table 43.1.1-1, Item 7. page 3/4 3-7 Reactor Protection Instrumentation), the main steam line radiatfon-high fiolation actuation instrumentation channel calibratioa P'able 4XX1-%, item 2.b).
and hgh system functional tests (TSa 4.3.14 aad 48~). Thc surve8laace interval would be extended. aa a cme-.
Ume basis, by 11 days ant0 the first cycle refueling outage scheduled to'egin cm September 15. 1967.
(3) TSs 4.3.2.2, 4.3.3.2, 43.3.3, 4.3.9.2, 45.1, 4.6.3.Z, 4.6.4.2, 4.6.68, 4.6.S.4, 4.6.5.5, 4.8.5.8, 4.7.1.1. 4.72, and 4.8.1.1 would be modified to extend the surveIHance interval, on a one-time basis, for Division I and Division H emetgency core cooling systems surveillance tests from August 1987 to the completion of the first refueling outage which is scheduled to begin on September 15, 1987. The affected surveillance tests are logic system functional tests and simulated automatic operation of all channels for the Isolation actuation instrumentation (TS 4XX2, Table 4~-
1, Items Ld and B.f); logh system functional tests and simulated automatic operation ofall channels for DivisionI trip system LPCI mode and LPCS systems (TS 4.3.3.2, Table 43.3.1-1. Iteta A.1), Division IItrip system LPCI B aad LPCI C systems (TS 4.3.3.2, Table 4.33.1-
- 1. Item B.1) and Divisioa I and IIloss of power (TS 4.3D.? Table 4%3.1-1. Item D.1); ECCS response time for each trip function (TS 4.3.3.3); logic system functional tests and aimuIated automatfc operation ofall channels for primary containment ventllatioa system - unit cooler A and B (IS 4.3.92 Table 4.3.L1-1, Items 1.a and I,c); system functioaal test for LPCS pump, LPCIA pump. LPCI.
B pump and LPCI C pump (TS 44 1)'ystem functional test of the containment unit coolers (TS 4,6.3&c);
verification that each automatic isolation valve actuates to its Isolation position on an isolation test signal (TS 4.6.42. Table 3.8.4-1); verification that each secondary Isolation system automatic isolation damper actuates to its Isolation system on a containment isolation test signal PS 4.B.K3.b, Table 3:6.5.3-1); system functional test ofthe standby gas treatment system and demonstration that the filtertrain starts and isolation dampers open on a simulated automatic initiation signal (TS 4.8.5.4, Items d.1.a and d.3.b); system functional test ofthe shield building annulus mixing subsystem an'd verification that the subsystem starts and isolation dampers open on a simulated automatic initiating signal (TS 4.8.S.S, Items b.1.a and b.3.bJ; system functional test of each fuel building ventilation charcoal filtration subsystem and verification that the subsystem starts and isolation dampers actuate correctly on a simulated automatic hltfetion sfgnal pS 4,6.5.6, Items e.1a and cL3.bJ; verfffcatfoa that each automa8c valve of the standby sen6m water scbsystetn actuates la the correct paaitfon aad the pamp starts on a nttrnud service water her pressure signal (TS 4.7.1.'E'b); verfficatfon that ecch mah control room afr ccmditfonfng subsystem automatically switches to the emergency mode ofoperation on a LOCA emergency mode actuatioa teqt signal and that the isolation valves close within 30 seconds and that the control room is maintained at a positive pressure (TS 42~~); and electrical power systems survelllances (TS 4.L1.%2, Item f).
Basis forproposed no significant koscrrds consideration determimrtionr The Commission has provided standards for determinhg whether a signlficant hazards consideratioa exists as stated h.10 CFR 5082(c). Aproposed amendment to an operating license fors facilityIav'olvcs no algnificant haiatcis cmakietatioa Ifoperatha ofthe facility in accordance with the proposed amendment would not: (1) hnrolve a signlficant hcrease ln 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)
In'volvo a significant rcduc6oa in a margin of safety. The IIcettsee adchessed the above three standards lttthe amendment application.
(1) Extension of the hterval for perfonnhg the functional test ofthe penetration valve leakage control system. The lhenscc stated the followingwith regard ta thc three stand archa
- a. %>> proposed change docs ncrt Irivcrtvca algnificant Iactcaae fn the ptobabIHty or araacqucncca of an acddcnt pteviouaiy evaluated because the increase In aurvcifiancc interval wiiinot result Itic reduction in system teIIsbiHty nor wHIit effect the abiHty ofthe system to perform Its dcsigrt frmction. This Ia demonstrated by the continued monthly and quarterly functionai and operational testing aa reqptlted by the tedmtcal apcclficathns es discussed above.
The system operation and design will therefore remain as described Itr the FSAR and as a result the response to an event wIII remain as anatyzcd.
- b. This change willnot create the poaaibiH> ofc new or different kind of accident from any acctdcnt ptcvtauaty cvaluatcd because It docs not Involve any changes to the system amfiguration or operation. A change in aurvciltanca Interval willnot create any new accidents.
- c. The proposed change wilInot algnificsntly reduce a metgtn ofsafety because the reliabilityof the system to petfmm Itifanctioii ia not atgnlficantly cffcctccL The system design, operation. and
.ability to hmctian when required remairi
Federal Reglgter' Vol. 62, No. 126 / Wednesday, July 1, 19N' Not(ceo, 24$53 I
unchanged. Additionally, as delineated ln the Justiflcatlon, the proposed change wfllnot affect the performance requirements ln the LimitingConditions of Operation contained In the Technical Speclflcatlon. Thus,'hi, margin of safety ls not Impacted.
(2) Extension of the Interval for ',
performing the main 'steam IItte radiation-high channel callbratione and
.'ogic system funcUonal tests. The licensee stated the followingwith regard to the three standards:
i
- a. No signifIcant hcrease in the probability or the consequences ofen accident previously evaluated results from this change because:
'he system design, function and configuration are not changed or affected.
The detector channels continue to be checked each shift. This check willIdentify any drift
'r degradation of the system. Driftassociated with the, detector would be insigntflcant when coinpared to the radiation levels associated with a failed fuel event. The logarithmic radiaUon monitor, which ls the primary source ofdriftw01 continue to receive the current 80 day surveillance testing. With the change in the surveillance interval for the MSLRH isolation actuation instrumentation. the loop remains fully operable and responds with the necessary accuracy to detect high radiation and initiate an isolation actuation. There is no signiflcant increase for the probabflity of failure of the tested components with the change to the surveillance interval, There Is no Increase in the probability or consequences of an accident previously evaluated because the system willcontinue to respond as designed.
- b. This change'ould not create the possibflity of a new or different kind of accident from any accident previously evaluated because:,
The design response of the instrumentation and the system remains the same and is unaffected. No change Is made in the design operation or configuration of the instrumentation therefore, previously analysis and evaluations remain valid. No credit is takeri for the MSLRH instrumentation in any design basis FSAR arialysis.
- c. This change would not involve a significant reduction in the margin of safety because:
The functional reiponse of the system will remain as designed during the extended surveillance period because of the redundant design and the reliabilityof the components.
As discussed In the justiflcation above, the detectors have been found to not drifta signiflcant amount during this extended suiveiflance period and willcontinue to receive functional testing each month of operation. A review of the equipment design, FSAR and SER cormnltments; and system performance requirements has confirmed the surveillance extension Is within the component capabflity and does not conflict with present systein requirements. Since tlie design and performance is not sensitive to the requested change and the response of the system willremain as designed and as described In the safety analysis report the margin of safety has not been signiflcantly reduced.
(3) DIvfsfoa I and IIemergency core cooling eyetem8 eurveIIltmce teste, The IIcensee stated the followingregard to the three standards a No slgniflcant hcrease In the probabflity, or the consequences of an accident revlously evaluated results from this chaiige causee Previous testhig recently conducted at this
'acflltyduring the preoperatlonal teat phase and during the power ascension test phase demonstrated Indepththe reliabflityand perfonnance capabfllty of the ECCS systems during various hltlating modes ofa LOP/
LOCAevent. In addiUon, an extensive survefllance program exists at this plant to demonstrate conUnued operability and performance ofthe ECCS systems required to mitigate the occurrence ofsuch an event. Due to the reliabflityproven by the previous testing and the conUnued proven operabflity obtained by frequent ongoing surveillance testing. extension of the 18 month ECCS surveiflances wfllnot result in a signtflcant increase In the probability or consequences of a LOP/LOCA event.
- b. This change would not create the
'osslbflity ofa new or different kind of accident from any accident previously evaluated because:
This change allows one time extension in the aflowed interval for which the surveillance is to be performed. This extension therefore. does not introduce a new mode ofoperation. Since no new or different kinds ofaccidents are introduced by extending the surveillance interval, then the possibility of creaUng an accident not previously evaluated does not exist.
- c. This change would not involve a significant reduction in the maigtn of safety because:
The deinonstrated reliability caused by,
'ecently conducted testing during the'tart-up phase of this plant, es well as extensive ongoing surveillance tesUng designed to determine operabflity and to measure performance of ECCS systems. ensures a margin of safety is maintained to offset the effects that would be caused Ifa LOP/LOCA event occiuTed. In addition, recent changes made to the plants protective tonal trip system resulting from the January 1988 LOP event. has reduced signiTicantly the possibility of a similar event happening again. ALOCA event, being classiTied as a limitingfault condition, maintains a probability of less than one tenth of one percent chance of occurring over the 40 year cycle of this plant. Leak detection systems and ISI inspecUon programs exist to detect
~ and prevent pipe and vessel failures.which could allow a LOCA event to occur.
~
Extending the performance date of the 18 month ECCS surveillance until after commencement of refueling would therefore, not impose a measurable reduction in the margin of safety.
The staff hae reviewed the licensee's no significant hazards consideration determination and agrees with the analysis.
Local Public Document Room
'Location: Government Documents Department, Louisiana State University, Baton Rouge, Louisiana 70803 Attarneyforllcenstta Troy B.Canner, Jr., Esq., Conner and Wettedude, 1747 pennsylvania Avenue, NW, Washington, DC 2000$
- NRCPraject Dlrttctar; Jose A. Calvo GulfStates UIQIthis Company, Docket No. 50458, RIver Bend.Statfaa, UnIt 1 West FeIIciana Pariah, Louhhna Date ofamendment requesti May 15, 1987.
Descript(on ofamendment reque'aL'he proposed amendment would revise the Technical Specifications (TSe) ta extend the surveillance Intervals, on a one-time basis. until the refueling outage scheduled ta begin on September 15, 1987, for the reactor vessel steam dome pressure'-high reactor protection system instrumentation and isolation actuation instrumentation, the main steam line flow-high Instrument loops, the primary containment, secondary containment, and reactor water cleanup level 2 and main steam hte leveL Isolation'actuation instrumentation, atld the automatic'depressurhatfon system (ADS) trip system reactor vessel water level-Iow level 3 and the, reactor core isolation cooling (RCIC) system reactor
'ater level-high level 8 actuation instrumentation. The proposed amendment would modify the TSs as follows:
(1) TSs 4.3.12, Table 4.3,1.1-1, 4.3.2.2, and Table 4.3%1-1 would be modified ta extend the interval for channel calibration and logic system functional test for the reactor vessel steam dome pressure-high reactor protection instrumentation and the reactor vessel pressure-high Isolation actuation inatrumentatlan. These surveillance tests are currently to be performed every 18 months. This one-time extension request ia for approximately 31 days until the forthcoming refueling outage.
(2) TS 4.3~ and Table 432 1-1 would be modified ta extend the channel calibration and logic system functional test surveillance intervals for the main steam line flow-high Instrument loops.
These surveillance tests are currently to be performed every 18 months, This one-time extension request ia for approximately 24 days (from August 21, 1987) until the forthcoming refueling
- outage, (3) TS 4.3.2.2 and Table 4,3,?.1-1 would be modified to extend the interval for channel calibration and logic system functional testing for the primary containment. secondary containment, and reactor water cleanup level 2 and main steam line level 1 isolation actuation instrumentation. These surveillance tests are currently to be
Federal Roghtes / VaL 5?
Ncx 2?6 / Wednesday, 7ttly 1, 296y I Notices performed every '18 mcmthL This ocM-time extension request is for apprmima tely 3b daya (fracn August 2$,
1985) until the forthcoming refuellns outage.
(4) TSs 4.3.3.2, Tatile 43.%1-2, 4.3.%2.
Table 4.3.5.1-2, end 4.5.2 wauld be modifled ta extend the ADS trip system reactor vessel water level-law'evel 3, and RCIC system reactor vessel water level-high level 8 actuation instrumentation surveiOance frequency.
These surveillance tests are currently to be performed every 18 manths. This one-time extension request is for approximately 21 days (f'rom August 25, 1987) until the forthcoming refueling outage.
Basis forprapased na significanI hazards aansiderarian detenninatian:
The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.M(c). A proposed amendment to an operating license for a facilityinvolves na significant hazards consideration ifoperation of the faciUty in accordance with the proposed amendment would not: (1) involve a significant increase in the probability or consequences ofan 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 signiflcant reduction in a margin of safety. The licensee addressed the above three standards in the emendment application.
(1) Reactor vessel steam dome pressure-high reactor protection instrumentation and reactor vessel pressure-high isolation actuation instrumentation surveillance extension.
With regard to the three standards. the licensee stated:
- e. No significant Increase In the probclity or the consequences of an accident previously evaluated results from this change because:
The system cfcsfgn, funcUon and configurcUan are not changccl Tbe requested extension may result in greater transmitter drift.This drift has been cahiilatcd to be 127 psig. The drift used In the FSAR analysis h 15 psig. Channel function h assured by surveillance ofchannel Instruments. This request willnot result'In any change In setpointe. allowable value or PSAR analysis.
In addition. a NPRDS search indIcated e very low probability for equipment failures with regard to manual switches and cuxiliery relays. Since there is no change Ia the bap's operation, or set points there le no slgnlficant
'increase In the probability oc consequence of previously evaluated accidents.
- b. This change would not create the possibility of a new or different kind of accident from any accident previously evaluated because:
This request wIU not result in ncw mades or configuraUons of plant operations. The design aacl opcraUon of the instriuncntatkm and the system remidne the same. thcrefcue.
previous analysis and evelaaUoas remahi valid.
- c. Tbia change woubl not involve a significant ceducUon in the margbi of safety baca ass:
Performance of the system end the components remain conshtecct with tbe requirements of the Technical SpcclficaUoas and FSAL Current setpoints allow for channel driftof up to 15 pstg. The cdudated driftIncluding this extenshn Is 12.7 pstg.
Therefore there h no change to the aUowcbh value or nominal trip setpoint and the mccgtn
. of safety Ls not reduced.
(2) Channel calibration and logic system functional (LSFf) test surveillance interval extension far the main steam line flow'-high instrmnent loops. With regard to the three standards, the licensee stated:
- a. No significant increase in the probability or the consequences ofan accidcat previously evaluated results from this change because:
The cxhting Tcchnhal Spcclficathn 'Mp Sctpoint and Allowable Value can accommodate an additional calculated drift for e 30 month channel cafibraUon interval Furthermore. the increased LSFf survcIUance Interval results in no significant probability of en MSIVholation logic faibue.
- b. Thh change would not create the possibility of a new or dtffcrent kind of ecctdent from any accident previously evaluated because:
This change does not dchte or reduce tbe functional capability, of the MSLHow-Higb instcumcntaUon. Therefore, no new kind of ecddent can result from this change and tbc response to an event willbe as analyzed.
- c. This change would not involve a signlficant reducUon In the margin ofsafety because:
The instrument sctpoint Is not'changed aa shoukl it change as a result of this sulvciUailcc extension Tbe cucrcn't sctpohlt has en afiowence for 5 psld of drift.Tha cakulctcd driftfor a 30 month (34 months plus twenty-Gvc percent) siuveifience interval is 1.41 pciilTherefore. the caladatcct driftis well within the aUowcnce fordriftas assumed in the analysh. The Intent of tbe Tcchnkal SpccBIccUon basis is mct because no significant ceducUan In the mergtu of safety or cffccUvencss of the h4SL Fiow49gh Instrumcntat tan hc mitigeUng tbe consequences of an MSLbreak outskle contabuncnt h Invobed.
(3) Channel calibration aud logic system functional testing eurveiflence extension for the primary containment, secondary containment, and reactor water cleanup level 2 and main steam
'line level 1 isolation actuation instrumentation. With regard ta the three standards, the licenme statech
- c. No stgntficant increase in the pcobebIBiy or the consequences ofan accident previously cvsbiatcd results frocn this change because:
It willnot result in a sfgnIIlcant reduction ia system ceHebIBty nar effec the ebIBty of the system to perform Its design functfon. The increased cefibraUon hitcrval does nat effect current instrmncut sctpoints duc to cxhUng designmmgtu. The system wiUcontinue to funcUan within the exhting design bases and analysis. The change in LSFI'urveillance interval Ia supported by successful oparathn of the InstrunlntaUon durmg stertup testing end inIUatoperation. In addition, NPRDS reports no manual switch failures and a MTBFot 239~ bours for auxiihry rdayia thus, showing bfgb rcBabifity.
- b. 'Ibis change would not create the posslbIUty of a new oc diffcrcnt khidof accident from any ecchhnt previously evaluated bccausc-There Is no change to system configucaUocc or analysts. The change in surveillance interval does not create any new types of acddcnh.
- c. Thh change would not involve a significant reduction in the margin of safety became:
The change in caBbration test interval does not bnpact Instrument sctpoints. The cckuiatcd drifthr within the ellowaMe value as given in the Tcchidad Specifications.
Current setpoints aHow forchannel driftof up to 4 Inches ofwater. Thc calcabitcd drift inchuBng this extension is 3M Inches. There Is no change to anelyUcel limitused In any enelysh. Dday In the logic system funcUonaI test docs nat significantl effect the probability of system failure. Therefore, this change does not significantly reduce the margin of safety.
(4) Extension ofthe surveiflance frequency for the ADS trip system reactor vessel water level-low level 3, and RCIC system reactor vessel water level-high level 8. With regard to the three stenchtrtls. the licensee statech
- a. No significent increase In tbe pcobebIBty ce tbe consequences of an accident previously evaluated results from this change because The requested change to the survctUance interval has been found ta be within the present design value for the eclpoint driftend willremain wtthbt tccbnkal specHica Uon allowable values for the requested extension.
The change is also found to have no significant effect on the system logh function because of the system design cnd tefiabfilty ofthe components. In addition. NPRDS reports show thc LSFT and auxiliary relays to be highly refieblc and not sensitive to test interval Therefore, there is no change to the current safety analysh rcquhed.
IxThis change would not create the possibifity ef a new or different Had of'ccident from any accident previously evaluated because:
The Increase In the reactor water level IaetrumcntaUon survcillence interval does not Increase'thc passibility ofan accident ar e mclhmcUon of e different type than previously evaluated since there is na change in funcUon or hardware.
- c. Tbts change would not Involve e signHIcsnt reduction In the miughi ofsafety because:
The change In this reactor vcssc? water lcvef histrmncnta Uan sucveillencc Interval does nat Involve c reduction In the margin of safety since the instruments setpoints cnd
Ftderal'oehter II VeL 52, No. 126 / We'daeaday, Jell 1, 1QSl / goQoea I
'llowable values ars not changed and the calculated drifts are well within the
~
allowable values. Since the change aielntalns the present safety analysis. there I~ no significant reduction to the margin of safety.
The staff has reviewed the licensee's no significant hazards Consideration determination and agrees with the analysis.
Local Public Document Boom Location: Government Documents Departinent. Louisiana Rate University, Baton Rouge, Louisiana 70803 Attorneyforlicensee: Troy B. Conner, Jr., Esq.. Gonne'r, and Wetterhahn, 174?
Pennsylvania Avenue, NW, Washingto'n, DC 20008 NBCProject Director. Jose A. Calvo Long Isla'nd Lighting Company, Docket No. 50-322, Shorehem Nudear Power '
Station, Suffolk County, New York Date ofamendment request: April24, 1987 Description ofamendment request:
These proposed, changes consist of modiifiications of certain emergency diesel generator license conditions as described in attachm'ent 3 to the Operating License NPF-38. These changes allow the licensee to Implement the staffs requirements which estabfish the basis for the continued qualification, reliability and operability of the TDI engines over'he entire life of the, plant,,
and involve detailed maintenance and surveillance activities for critical engine components. These proposed license changes incorporate all of the TDI Owner's Group or Pacific Northwest Laboratory (PNL) recomniendations, which were reviewed and approved by the Staff in NUREG-1218; Basis forproposed na significant
'azards consideration determinationt TheCommission 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
'acility involves no significant hazards co'nsideration ifoperatfon of th'e facility in accordance with'the proposed amendment would not (1) Involve a
~
significant increase in the probability or consequences of an,acddent previously evaluated: (2) create the possibility of a new or different kind of accident from any acddent previously evaluated; or (3) in'valve a significant reduction in a margin of safety.
The licensee has determined. and the, NRC staff agrees. that the proposed amendment willnot:
(1) involve a significant Increase In the probability or consequences of an accident previously evaluated. The prosed license
'onditions as mandated through the StsFs Final SER (NUREQ-1216).'stablish the basis for the continued design adequacy sad qualification of the SNPS TDIengines for nuclear standby service, Iis rsqufredby GDC 17 ofAppendix h to 10 CFR Part M. The Incorporation of the moat critical peifaBa maintenance/surveillance inspections for.
certain Phise I engine components as )lcsase coadftfoai willensure coatfaaid reliability, aad operability of these units over the lifed the plant. These proposed Ifcease conditions do aot involve any changes ia the systsai's design aad operability requirements aor ia the pleat operating conditions or parameters.'2) create the possibility ofan eccideat that Is different than alrisdy evaluated ia thii USAR. The proposed license conditions involve maintenance aad surveillance activities for critical eagfae compoaeats ia order to easuri'proper aad safe operatfori of'he engines and mefntaia the adequacy aad availability of these units for audear standby service. These proposed license condition amendments do aot affect any plant conditions or parameters aor do they alter the system design'aad operability requirements. Therefore, the possibility of a n'w or different kind of accident from any accident previously evaluated caaa'ot be created. (3) Involve a significant reduction ia the margin of safety as de'fined ia the bases to Tediaical Specifications 3/4&,hllrequired inspections for critical erigine components that are Incoiporated In these proposed license coadftfoai meet or exceed those recommended by the vendor and the TDI EGG Owner's Group DR/QR program.'These inspections do not alter the functional safety requirenieats of,the emergency dfeiel generators. thus there is no impact ori the results'aad conclusions of the USAR Chapter 15 analyses sad ao impact on the margin of safe'ty. Also, all requfreiuenti of the existing license'conditions as Imposed by the NRC ere maintained until startup fiom the first, refuelfag outage. Furthermore, all proposed new license conditions forpost first fuel.
cycle have beeii revfewe'd aad, approv'ed by',
the NRC "staff and 'documented In NUREG- "
1218. Therefore, the proposed license changes do aot Involve a sigaiffcant reduction ia the margin of s'efety.
Accordingly, the Commission proposes, to determine that the proposed changes to the )fcense involve no significant hazards considerations, Local Public Document Boom-location: Shoreham-Wading River Public Library. Route 25A, Shoreham, New ',
York 11788 Attorneyforlicensee: W. Taylor Reveley, IILEsq., Hunton and Willianui, P. O. Box 1535. Richmond. Virginia23212 NBCProject Director: Walter R Butler Northern States Power Company, Docket No. 50.283,'Montfce)lo Nudear Generating Plarit, Wright County, Mnn'esota
'ate ofamendment request: February 4, 1987.
Description ofamendment request:
The application for amendment proposes new Rod Block Monitor (RBM) setpoints as a result ofMonticello specifi analyses performed by General Electric. The setpoints replace existing RBM setpolnts that were determined from a generic analyslL Table %24 of the Tec)mica) Spedficat(ons would be
',c)fanged to reflect the new RBM Upscale Setpolnts, The minimum critical power ratio (MCPR) would be changed in the associated bases.
Basis forproposed no st'gnificant, hazards consideration determinatiarL'he Commission has provided stander'ds for determining whether a sfgnffica'nt hazards consideration exists (10 CFR 50.92(c)). A proposed amendment to an operating Bcense for.a facilityinvohres no sfgnfficant hazards
'onsideration if'operation of the facility in accordance with the proposed amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated: or (2) create the possibility of a new or different kind of acddent from any accident previously evaluatedor (3) involve a significant reduction in a margin of safety.
The )fcensee has evaluated the proposed RBM setpoint changes and has conduded that no significant hazards consideration exists because of the followingreasons:,
'eneral Electric has performed an,'nalysis supporting the proposed setpoints for an initialMCPR of1M.
This analysis'used the same methodology used to calculate the existing setpoints. The only change is a differ'ent value of initialMCPR
', 1.30 instead of 120. Since the current cyde's lowestallowed MCPR Is 1.38 and the analysis allows setpoints,actually slightly higher, than those p'roposed, the proposed setpoints are conservative.
Therefor'e, the proposed setpoints willr not involve a significant increase in the probability, or consequences of any accident previously evaluated.
The proposed change in RBM setpoints willnot create the possfbifity of a new or different kind of accident
~
since the change makes relatively minor diariges to the existing setpoints.
Since the proposed setpoints have been determined by approved NRC methodology. this change willnot involve a significant reduction in the margin of safety.
The staff has r'eviewed the licensee's evaluation'and agrees with their conclusions. Accordingly, the staff proposes to determine that th' requested action does not involve a significant hazards consideratian.
Loca/Public Documeri t Boom location: MfnneapoUs Public Library, Technology and Sdeace Department,
1
\\
- 24556, Federal Resister / Vol, 52, No. 128 / Wednesday,'July 1, 1887 / Notices 300 Nicollet Mall, Minneapolis, Minnesbta 55401.
'ttorney forlicensee: Gerald Charnoff, Esq., Shaw, Pittman. Potts and Trowbridge. 2300 N Street, NW.
Washington. DC 20037.
NRCProject Director. David L Wigginton. Acting.
Northern States Power Company, Docket No. 50-263, MonflceHo Nuc)ear Generating'Plant, Wright County, Minnesota Date ofamendment request: February 18, 1987.
Description ofomendment request:
The proposed changes would c)arify the Technical Specification requirements for IRM and APRM scram instrumentation operability by revising Table 3,1.1 to eliminate the requirement for IRM operability while in the Run Mode and to delete the APRM downscale scram, Table 3.1.1 note 2 would be changed to read: "For an IRM channel to be considered operable, its detector shall be fullyinserted." Note c would be deleted; and other clarifying changes would be made.
Basis forproposed no significant hazards consi deratian determinati ant The Commission has provided standards for determining whether a signiflcant 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 significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibifity of e new or different kind of accident from eny accident previously evaluated; or (3) involve a significant reduction in a margin of safety.
A literal interpretation of the affected sections now requires placing the plant in a "half-scram" condition to perform required testing and,maintenance.
This makes the plant more susceptable to spurious trips and initiation of safeguards equipment.
The proposed changes would clarify the intent of the original specification by clearly definirg the scram functions needed to be operable in each mode of operation. The allowable bypasses assure that the single failure criteria are satisfied for the required scram function of the IRMs and APRMs. These changes do not involve modifications of the reactor protection system wiring or circuitry thus, by design, overlap between the )Rlvfs and APRMs is assured. Therefore, the requested changes willnot involve a significant increase in the probability or consequence ofany accident previously evaluated. For the same reasons, the proposed changes willnot create the possibility of a new or different kind of accident, nor willthey involve a signflicant reduction in the margin of safety.
Based on the above considerations, the Commission proposes to determine that the proposed changes do not involve a significant hazards consideration.
Local Public Document Boom location: Minneapolis Public Library, Technology and Science Department, 300 Nicoflet Mall.Minneapolis, Minnesota 55401.
Attorneyforlicensee: Gerald Charnoff, Esq., Shaw', Pittman. Potts and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Project Director. David L Wigginton, Acting.
Philadelphia Electri Comp'any, PubHc Service Electric and Gas Company, Delmarva Power and Light Company, and Atlantic City Electric Company, Dockets Nos. 50-277 and 50-278, Poach Bottom Atomic Power Station, Units Nos. 2 and 3, York County, Pennsylvania Dote ofapplication for amendments:
January 21, 1987 Description ofomendment request'he proposed amendments would change the Technical Specifications (TSs) for the Peach Bottom Atomic Power Station, Units 2 and 3 by (a) providing an extension of 30 days in the current 60 day requirement ofTS 8.9.2.h(2) on Page 259 for filingthe Semiannual Effluent Release Report, and (b) eliminate a reporting requirement in TS 6.92.a on page 257 of loss of shutdown margin that is redundant to the Licensee Event Report (LER) requirements.
The current TS 6.9.2.h(2) requires that the Semiannual Effluent Release Report covering the previous six months of operation shall be submitted within 60 days after January 1 and July 1 of each year. The licensee states that the extension is appropriate since the amount of time required to complete the
'steps in preparing the report is not sufficient to allow adequate review of the report prior to the 60 day deadHne.
Consequently, the licensee encounters the need to file followup reports for corrections and to supplement the original report. The licensee states that the additional 30 days to submit the Semiannual Effluent Release Report will ensure that an adequate period of time is available to send the effluent samples to a vendor for analysis, receive and review the data and prepare a complete report. Additionally. this willprovide a reasonable amount of time forreview by the Hcensee's technical staffs and management, and to identify and correct errors.
The current TS 6.92.a on page 257
.requires that a gpecial Report be submitted to the NRC for loss of shutdown margin. The licensee requests that the reporting requirement in Technical Specification 8Al&abe deleted since it is redundant, except for the reporting schedule, to the new LER Rule. The Hcensee further states that loss ofshutdown margin is reportable under the provisions of the LER Rule (10 CFR 50.73(a)(2)(i)(B)). Consequently, the reporting requirement ofTechnical Specification 6.92.a is redundant to the reporting requirements of 10 CFR 50.73.
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 50.92(c). A proposed amendment to an operating license involves no significant hazards considerations 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
'ova)uated, or (2) create the possibility of a new or different kindof accident from any accident previously evaluated or (3) involve a significant. reduction in a margin of safety.
The licensee has evaluated the proposed amendment to TS 8.92.h(2) against standards in 10 CFR 50.92 and has determined the following:
Operation of Peach Bottom Units 2 and 3 in accordance with this change would not:
(1) Involve a signiflcant increese In the probability or consequences o! an accident previously evaluated because the collection and analysis ofroutine plant data, and the preparation time to transmit the Information to the NRC Is independent of plant design and operational characteristics that can impact potential accidentiui (2) create the possibility of a new or different kind of accident from any previously analyzed because the routine, collection and transmittal of data does not eetabflsh a potential new acctdent precursor.
(3) involve a slgniflcant reduction In the margin of sefety because the addit)anal 30 days to submit thts routine plant data does not impact the health and safety of the public since signtflcant increases In radiological effluent releases that may be indicative of an inherent defect in plant design or operetion.
are reported promptly under the provision of 10 CFR 50.72 and 10 CFR 50.73 and are therefore not impacted by this amendment request.-
The licensee has evaluated the
, proposed amendment to TS 6.92.a
Fedaca1 Reghtfer / VOL I, No, 1t8 / Wednesday, paly 1, lNF ] Xesoee
?ASSF gslnat the standards in 10 CFR 5tLSfE goal hss determined the foHowfnfp
'paratfoii of Peach Bottom Units 2 amf 3 h tcmzgance with thia change would uou (tl involve a affpifficsntincrease in the pohsblllty or consequences ofan accident pravfouafy cafcuhted becauae the chsiigs goes iiot gfmhfah repo requirements but merely eliminates a redundaat reportfiig zaqiilrement for Loss ofSlmtdown Msitffic
[2) create the poaaiblffty ofa new or gffferent kind ofaccfdent from any pravfouafy analyzed because reportfiig zaqigzementa do not eatsbliah a potential.new
, acdgeiit pmcuraon (3) hvolva a af8niffcant mlactfou fn a
-maigfiiofsafety becauae the cunent reporUng requfrementa in 10 CFR QL72 aiid 10 CFft 50.73 covers alf events, IncfudfiigLoia of Stiutdown Maisin, that may impact the margfn ofaafety. and require the NRC to be notified.
Based on the above reasoning, the Hcenaee has determined that the proposed amendment does not involve a significant hazards consideration. The NRC staff has reviewed the licensee's no significant hazards consideration determination and agrees with the licensee's analysis. Based on this review, the staff, therefore, proposes to determine that the requested amendment does not involve a significant hazards cansideration.
Local Public Document Room location: Government Publications Section. State Library ofPennsylvania.
Education Building. Commonwealth and Walnut Streets, Hamsburg, Pennsylvania 17128 AltonieyforLicensee: Troy B. Canner, Jr., 1747 Pennsylvania Avenue, NW.
Washington. DC 20008 NRC Pmj ect Director. Walter R Butler Public Service Electric gi Gas Company, Docket No. 5MS4, Hepe Creek Generating Station, Salem County, New Jersey Date ofamendment request: April30, 1987 Description ofamendment requeat:
The amendment would mddify Technical Speciflcation Figure 3.2.38 to increase the minimum critical power ratio (MCPR) versus tau at rated flow.
The current Technical Specfflcatfon provides a single MCPR limitof 1.20 for eH values of tau from 0 to 1,0 at rated flow. The proposed change would increase the MCPR Bnearly from a lhnit of 1.20 for tau equal to zero to a limitof 1.23 for tau equal to 1.0. This change to increase the MCPR limitis in the conservative direction and is more restrictive than the current limit.
Boai a forproposed na aigjiificant hazards consideration'determinatiozu The Commission has provided guidance concerning the application af its standards set fozqh fn 10 CFR 5088 by providing certain examples (51 PR 7751).
One ofthe examples, (li),ofan amendment likelyto involve na significant hazards consideration relates to "Achange that constitutes aa additional ffmftatfon, restriction, ar control not presently fnohzded in the technical specfflcatfons, e.g., a more stringent suzvefflance requirement." The proposed amendment relates to this example because itwould impose more restrictive limits on the aHowed MCPR for any value oftau above tau equal.to zero.
Therefore, the Cannniseion proposes to determine that the proposed amendment fnvolves no sfgnfflcant hazards considerations.
Local Public Document Boom location: Pennsvifle PubHc Hbrary, 190 S.,
Broadway, Pennsville, New Jersey 08078 Attonieyforlicensee: Troy B. Conner, Jr., Esquire, Conner and Wetterhahn, 1747 Pennsylvania Avenue, NW; Washington. DC 20006 NRC Project Director.. Walter R.
Butler Public Service Electric tk Gas Caznpeny, Docket No. SM54, Hope Creek Generating Station, Salem County, New Jersey Date ofamendment request: May 1, 1SS7 Deaan'ptian ofamendment requeat:
The proposed amendment would revise the numbering ofcurrent Technical Speciflcation Sections 3.7.8 through 3.'/.10 and ofTables 3:/.9-1 and 3.7.10-1 by renumberfng them Sections 3.78 through 3.7.11 and Tables 3.7.10-1 and 3.'/.11-1 respectively. In addition, the section numbers referenced in the text of these sections and in the text of Section 11KB willbe changed to, correspond with the appropriate new section numbers. These numbering changes are proposed in order that the Technical Specfflcatfon section numbers willbe consistent with the numbers referenced in the Hope Creek Generating Station proceduzee. Anew Sectian 3.7.8.wiH be added which wiHbe blank except for a note to indicate that the purpose of the section is to maintain numerical continuity ofthe section numbers.
Basis forproposed no significant hazards consideration determinati ozu The Commission has provided guidance
'oncerning the appBcatfon of its standards set forth in 10 CFR 5tL92 by providing certain examples (51 FR 7744).
One of the examples, (i), ofan amendment Bkely to involve no significant hazards conslderatian seletes
~
to "Apurely administrative change to technical specfflcatfons: forexample. e chazzge te achieve cezzsfstency throughaut the taohdcal specfffcatfaus, correctfonefen eztor,era change in nomenc)eture." The proposed amendment es dhcussed above ze)ates to this exemple because,ft anly changes 81e section Izumbers.
Vhe6gfozvz, the Commission proposes to determine that the proposed amendment involves no sfgnfflcant hazards consfderatf one.
LocalPublic Document Roam locatiaru PennsvfHe Publfc library, 190 S.
Broadway, PennsviHe, New Jersey 08070 Attorneyjbrlicensee: Troy S. Canner.
Jr., Esqufre, Conner and Wetterhahn, 1747 Pennsylvania Avenue, N%f.
Washington, DC 20008 NRCPro/ect Director. Welter R.
Butler PubHc Service Electric h Gas Caaagaazzy, Docket No, 5M54, Hope Creek Generating Station, Safean Couzaty, Naw, Jersey Date ofamendment request: May 22.
1987 Description ofamendment request
'IIze amendment would modify the Technical Specfflcatfons ta peznzft temporary adjustment of the setpaints for the Main Steam Line Radiation-High.
High trip funotion. 'Ihe change would perznft tbe normal fallpower background radiation level associated with the Main Steam Lhe Radiation scram and fsolatfon setpaints to be increased during hydrogen fnjectfon testing. The puzpose ofthe hydrogen fnjectfon test is to determine the feasibility ofhydrogen water chemistry control es a means af reduchg intergranular stress corrosion cracking (IGSCC) af stainless steel piping. The setpoint increases are needed to compensate for anticipated increases in the main steam line radiation level during hydrogen injection. Tbe background radiation level increase during hydrogen injection is caused by higher levels of short half-Bfe N-18 carryover into the steam line.
The proposed modfflcatfon would allow this temporary adjustment to the setpoints to be made only when above 22 percent of rated power and would require that it be made within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to planned start ofhydrogen injection. It would also require that normal setpofnts be established within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ofreestabflshfng normal radiation levels after completion of the hydrogen injection and ta establishing power levels below 22 percent rated ye>ver.
A sbnffar change~ eppaoved for the pazyose af.hydrogen fnjectfonteitts at the MwfnIiRatch Nuclear Ident, Unit
24558 Federal Register / Vol. 52, No. 126 / Wednesday, July 1, 1987, / Notices 1 by Amendment No. 125 to the Hatch license, dated May 21, 1986.
Basis forproposed no significant hazards consideration determinationr The Commission has provided standards for detenpining whether a significant hazards coinijderatlon exists (10 CFR 50.92(c)). A proposed amendment to an operating license for a facilityinvolves no significant hazards consideration ifoperation of the fadlity in accordance with the proposed
amendment would not: (1) involve a significant increase in the probability or consequences ofan 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 ofsafety.
The licensee provided the following evaluation with its May 22, 1987 amendment request:
The proposed changes to the HCGS Technical Spediicatioas:
(a) Do aot Involve a sigaiflcaat increase ia the probability or consequences ofan accident previously evaluate*The only design basis acddeat which takes credit for the main steam )Inc high radiation scram aad
!solatioa setpoiat Is the Control Rod Drop Accident (CRDA) as described In FSAR Sectfon 15.49. Speciflcaliy. the Main Steam Isolation Valves (MSIVs) are assumed to receive an LCR automatic dosure sfgaal at 0.5 seconds after detectioa ofhigh radiatioa h the main steam lines aad to be fullydosed at 5 seconds from the receipt of the dosure etgaaL The main steam Ifae radistioa monitors are provided to detect a gross failure of the fuel daddtag. When high radiation Is detected. a trip h Initiated to reduce the continued failure offuel daddiag.
At the same time, the mefa steam isolation valves are closed to limitthe release of fission products. The trip setting ls high enough above background radiation levels to prevent spurious trips yet low enough to promptly detect gross faflures ia the fuel daddfag.
As Indicated In the NEDO Report (Reference 1), the consequences of the CRDA are most severe under Hot Standby conditions. In fact, the consequences of the CRDA are increasingly less severe above 10 percent due to a faster Doppler response aad a lower rodworih. Most Importantly, above 20 percent power, the consequences of the CRDA are mlatmsL Since the Main Steam Line Radiattoa Moaitor setpoiat willoa)y be adjusted for the purposes of the hydrogen Injection test at power levels above 22 peru.at. there Is ao sigaiftcaat impact on the probability or consequences of the CRDA.
Therefore, the change to ihe footnotes Ia the referenced TS tables have ao affect on the
'probability or consequences of an accident previously evaluated.
(b) Does aot create the possibility ofa new or different kind of acddeat from aay accident previously evaluated. The proposed changes do aot affect the design of any sefeiy.related systems aad as such do aot affect the performance of any safety
. fuactioas. The proposed changes do permit the performance of a hydrogen injection test; however, this test does not Introduce a new kind of accident since the presence of hydrogen Ia the primary system has already been analyzed (see FSAR Secttoa 82.5, 10.44
~
aad 1142.1) aad Is already iaoaitored aad controlled (see TS 3/4.3:/.10 aad 3/4.11>e, respsctlvsly). Further, ad discussed Ia Paragraph V above, additional protective measures are being applied which assure that the physical presence of test equipment does'ot create the potential for a different kind of accident to occur.
Since the TS changes themselves do not affect exisiiag system function, aor do they create a situation which has aot been previously analyzed aad appropriately designed for, Ihe change io the TS footnotes shown In Attachment 2 do aot create new or different accidents than previously evaluated.
(c) Does aot involve a'sigiuflcaat reduction
~
In a mergta of safety. The proposed temporary Increase In the Main Steam Line Radiatloa - High - High scram aad Isolation setpoints willbe penntsslble'only when reactor thermal power Is above 22 percent.
As discussed In Paragraph VLa above, the only desiga basis accident which takes credit for ibis scram aad holatioa trip function Is the CRDA. However, above 20 percent power, the consequences ofa CRDA are so minimal that they may be considered negligible (Ref. 1). aad hence, the change In the TS setpoiats has ao significant effect on the margins ofsafety for this acddent
'cenario.
The proposed change h necessary to conduct a hydrogen injection test which wfll Increase the carryover ofN-18. This ia turn, willcause the background radiatioa levels in the main steam system io be Increased. As discussed In Paragraph IVabove, several precauttoaary aad preplanning measures are being taken to maintain plant personnel exposures~ In addition, radiattoa levels willbe monitored to assure measured radiation levels are within acceptable site ALARA.Due to the relatively short half-IIfe ofN-10 (approximately 7 seconds), gaseous effluent release rates willaot be significantly affected. Therefore. It can be concluded that the proposed change willaot present a risk to the public health aad safety aor sigaiflcaatly reduce a margin of safety forplant personneL Baaed upon the discusstoas Ia the above three subparagraphs, PSEaG concludes that the proposed change does aot Iavolve a sigatflcaatly safety hazard.
Reference 1 referred to above by the licensee Is NEDO-10527, Supplement 1, "General Electric Rod Drop Accident Analysis for Large BoilingWater Reactors" dated July 1972.
The staff agrees with the licensees evaluation and conclusion as stated above. Accordingly, the staff proposes to determine that the requested amendment does not involve a signjficant hazards consideration.
Local Public Document Room
/ocation: Pennsvllle Public library, 190 S.
Broadway, Pennsville, New Jersey 08070 Attorneyforlicensee: Troy B. Conner, JrEsquire, Conner and Wetterhahn, 1747 Pennsylvania Avenue, NW, Washington, DC 20006 NRC Pmj ect Director: Walter R.
Butler Public Service Electric 5 Gas Company, Docket No. M-272, Salem Nudear Generating Station, Unit No. 1, Salem County, New Jersey Date ofamendment request: October 3, 1986 Description ofamendment request:
The proposed change would revise Technical Specification 42.2.2.e and Basis 3/42.2. The change replaces the F~ limits with statements referring to
'he Radial Peaking Factor LimitReport, which provides for a cyc)e-by-cyde determination of the F~ (Z) limits without the need to submit F~ Technical Specification changes. The F~ limitis a review criterion which is used to verify that the design neutronic calculations associated with the Reload Safety Evaluation are conservative. The Radial Peaking Factor LimitReport allows for cycle-by-cyde changes In the F
- limit, which refiect the variatlons to be expected based on the design calculations.
Asimilar change request forUnit 2 was submitted as LCR 8342, dated January 31, 1983, and was approved by'he Commission in Amendment 19, dated May 6. 1983, The proposed change makes the Unit 1 specification identical to that of Unit 2, and brings it into conformance with the Westinghouse Standard Technical SpeciTications (NUREG4452, Revision 4).
Basis forproposed no significan!
hazards consideration derermination:
The licensee provided the following slgnificant hazards evaluation per 10 CFR 50.92.
The proposed change to the Technical Spectficattoas Is administrative in nature in that It Is being made to achieve consistency to a previously approved change. The change does aot Involve a sigaiflcaat hazards coaslderatioa because operation In accordance with this change would not:
(1) Involve a sigatflcaat increase fa the probability or consequences ofan acddeat previously evaluated. The change involves only the survefllaace ofP~ as a veriflcatioa of the design models aad only changes the method ofdocumeatiag P~ limits.
(2) Create the possibfliiy ofa new or different kind of acddeat Erom any previously analyze*There are ao equipment.
Instrument, or setpotat changes related to the proposed change to Technical SpecUicattona (3) Involve a slgalflcant reduction In a margin ofsafety. The actual margin of safety as deflaed Ia the basis for the Fq technical specilicatioa rematas unchanged. ilnce the Fq limitI's unchanged. The radial peaking factor, F~(z) Is measured periodically to provtde
"~
~
~
Federal ReS(ster / Vol. 52, No, 128 / Wednesday, July 1, 1987 / NoUces
'4SSS Itgutsiice that the hot channel fector Fq(z) remains within Its limit.
The Commission has provided guidance concerning the application of the standards for determining whethe~ a significant hazards consideration exists (51 FR 7744, dated Ma'rcli 6, 1986), The proposed change corresponds to Example (I) for purely administrative changes to achieve consistency.
Therefore, on the basis of the licensee's evaluation. with which we agree. and because the proposed change corresponds to Example (i), noted above. the Commission proposes to determine that the amendment does not involve a significant hazards consideration, Local Public Document Room locatiom Salem Free Public library, 112 West Broadway, Salem. New Jersey
.08079
'Attorneyforlicensee: Mark J.
Wet terhahn, Esquire, Conner and Wetterhahn, Suite 1050. 1747 Pennsylvania Avenue, NW, Washington, DC 20006 NRC Pleat Director: Walter R.
Butler Public Service Electric si Gas Company, Docket No. 50-272, Salem Nuclear Generating Station, Unit No. 1, Salem County, New Jersey Date ofamendment requestt April20, 1987 Description ofamendment request:
, The proposed change requests modification ofFacility Operating License DPR-70 to incorporate (the Facility Attachment No, 13, dated October 1. 1986 to the US/
IAEASafeguards Agreement) along with clarifications as identified in (Salem Nuclear Generating Station, Unit 1 (SNG S1)
IAEASafeguards License Conditions, Revised) into the license.
Basis forproposed na significant hazards consideration determination:
The licensee has determined that the proposed change Involveb no significant hazards consideration under the provisions of 10 CFR Part 50.92, Based on the licensee's evaluation, we conclude that the proposed change is administrative in nature and as such conforms to Example (i) of (51 FR 7744).
The proposed amendment implements an IAEASafeguards inspection program and does not in any way affect the design bases or operation of the facility.
The purpose of the IAEAsafeguards inspection is to permit the IAEAto verify that special fissionable material at the facilityis not withdrawn (except as provided in the US/IAEASafeguards Agreement) from the'facility while such material is being safeguarded under the agreement. As such, the proposed amendment would not: (1) involve a slgnificant increase in the probability or consequence of an accident previously evaluated; or (2) create the probability of a new'or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. Therefore, the proposed amendment does not constitute a significant hazards consideration.
Accordingly. on the basis of the licensee's analysis, withwhich we agree, and because the circumstances seem to fitExample (i) above, the Commission proposes to determine that the amendment willnot involve a significant hazards consideration.
Local Public Document Room location: Salem Free Public library, 112 West Broadway, Salem, New Jersey 08079 Attorneyforlicensee: Mark J.
Wetterhahn, Esquire. Conner and Wetterhahn, Suite 1050, 1747 Pennsylvania Avenue. NW, Washington, DC 20008 NRC Project Dt'rector: Walter R.
Butler Tennessee Valley Authority, Docket Nos, 5M27 and 50-328, Sequoyah
'uclear Plant, Units 1 and 2, Hamilton County, Tennessee Date ofamendment requests:
December 17, 1986 (TS 76)
Description ofamendment requests:
Tennessee Valley Authority proposes to modify the Sequoyah Nuclear Plant Units 1 and 2 Technical Specifications to revise the containment spray response time, item 7.a ofTable 3.3-5, 'Engineered Safety Features Response Times.'he amendments would change the containment spray response time, for diesel generator loading. from 58 seconds to 208 seconds. This notice supersedes a previous notice dated February 11, 1987 (52 FR 4420).
Basis forproposed na significant hazards consideration 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 at the time a licensee requests an amendment, it must provide to the Commission its analyses using the standards in Section 50.92, about the issue of no significant hazards consideration. Therefore, in accordance with 10 CFR 50.91 and 10 CFR 50.92, the licensee has performed and provided the followinganalysis.
(t) ls the probability ofon occurrence or the consequences ofan accident previously evaluated in the safety analysis report significantlyincruased7 No, Ths containment spray (CS) system ls part of the Containment Heat Removal System. The primary design of the Containment Heat Removal Spray Systems is to spray cold water Into the containmsnt atmosphere when appropriate in the event of a losy ofcoolant accident (LOCA) and thereby ensure that containment pressure does not exceed the containment shell design pressure of12 psig. Temperature and pressure transtents shown In the Final Safety Analysis Report (FSAR) for the worst case LOCAremain relatively constant until switchover to sump recirculation. At this point. containment temperature drops sharply because the cold air exiting the fce condenser is no longer being warmed by the spray water. The temperature recovers and stabilizes at the outlet temperature of the spray heat exchangers once the spray pumps have been resterte*The temperature increases further after ice bed meltout. bui remains within design limits. Substantial margin still exists between the time sump recirculation begins and ice bed meltout occurs. Delaying actuation ofCS io 208 seconds willnot have any signiflcant effect on the containment temperetures and pressures as previously evaluated in the FSAR.
(2) ls the possibilityforan accident ofo new or diffen.nt type than evaluated previouslyin the safety analysis report created7 No. The proposed change makes a minor modification to an accident previously evaluated in the FSAILThe only change being made is the system response time. The evaluation ofthe proposed delay for CS actuation concludes that system operation end performance willbe as presently expected in matntainliig containment pressure and temperature design limits.
(3) ls the matin forsafety significantly teduced7 No. The evaluation bf the proposed delay for CS actuation concludes that containment pressure mitigation willremain within the safety limits.The delay for CS actuation will also delay the time before switchover to containment sump recirculation occurs.
However. adequate safety margin still exists between rectrculation flowactuation and ice depletion.
The staff has reviewed the licensee's no significant hazards consideration determination and agrees with the licensee's analysis. Therefore, the staff proposes to determine that the application for amendment involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County Library. 1001 Broad Street. Chattanooga, Tennessee 37402.
Attorneyforlicensee: General Counsel, Tennessee Valley Authority, 400 West Summit HillDrive, E11 B33, Knoxville,Tennessee 37902.
NRC Assistant Director: John A.
Zwolinskl
24560 Federal Register / Vol. 52, No. 126 / Wednesday, July 1. 1987 / Notices Tennessee Valley Authority, Docket Nos. SM27 and 5M28, Sequoyeh Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee Dote ofamendment requests: April14, 1987 (TS 87%2)
Description ofamendment requests:
Tennessee Valley Authority (TVA) proposed to modify the Sequoyah Nuclear Plant, Units 1 and 2 Technical Specifications (TS) to delete the functional test surveillance requirement (SR) 4.3.3.9 for testing automatic isolation of the release pathway from the steam generator blowdown (R-90-120 and R-90-121) and condensate demineralizer (R~225) liquid eff)uent radiation monitors due to instrument downscale failure. Also proposed is the addition of a functional test to SR 4.3.3.9 to demonstrate that e control room
, alarm annunciation results should these instruments incur downscale failure.
The radiation monitors in question
- vere neither designed nor intended to initiate an automatic isolation of the release pathway from an instrument downscale failure. Automatic isolation
. of the release pathway and. control room alarm annunciation Is required, and the TS remain unchanged, for an indication of measured levels above the alarm/trip setpoint or a circuit failure. The proposed change only affects the requirement for automatic isolation of the release pathway due to an instrument downscale failure. Control room annunication remains a requirement, as proposed. for an instrument downscale failure.
Basis forproposed no significant hazards consideration determinatiom The Commission has provided standards for determining whether a significant hazards determination exists as stated In 10 CFR 5082 (c). 10 CFR 60.91 requires that at the time a licensee requests an amendment, lt must provide to the Commission Its analyses, using the standards in Section 60.92, about the Issue of no significant hazards consideration. Therefore, in accordance with 10 CFR 60.91 and 10 CFR 50.92, the licensee has performed and provided following analysis.
(t) Does the proposed amendment involve a significant /rrcrease ln the probabll/ty or consequencee ofan accident previously evo/uated7 No. Thrt proposed amendment does not result h a change to the current plant configuration: rather. thc proposed amendment coirccts and clarifies a, au~eillancc requirement for hardware as currently Installed In the plant. Thus, the proposed technical spccification change involves no significant Increase In the probability or consequences of an acctdcnt that has been previously evaluated.
/2/ Docs the proposed amendmenl create rrrcpossibi/ity ofa new or differcrrt kind of accident from any accident previously eva/uated7 No. The proposed amendment Is not a result ofchanges In plant hardware, nor does It affect operating Ihitts. normal operating procedures. or emergency operating Instructtons for the plant. Thus, the proposed tcchnical spcciftcathn change does not create the posstbi0ty of a new or different kind of ccddent from any acctdent previously evaluated.
(3) Does the proposed amendment involve a significant reduction in a margin ofaofety7 No. The proposed amendment corrects and clarifies a surveillance requirement for hardware currently installed in the plant, thereby eliminating possible confiislon with performing that surveillance requirement. The proposed technical specification, therefore, does not Involve a significant reduction in a margin of safety.
The staff has reviewed the licensee's no significant hazards consideration determination and agrees with the licensee's analysis. Therefore, the staff proposes to determine that the application for amendments involves no significant hazards consideration.
LocalPublic Document Room location: Chattanooga-Hami) ton County Library. 1001 Broad Street. Chattanooga, Tennessee 37402.
Attorneyforlicensee: General Counsel, Tennessee Valley Authority, 400 West Summit HillDrive, E11 B33, Knoxville,Tennessee 37902.
'RC Assistant Director. John A.
Zwolinskl Tennessee Valley Authority, Docket Nos, SM27 and 5M28, Sequoyah Nuc)ear Plant, Units 1 and 2, Hamilton County, Tennessee Date ofamendment requests: May 15.
1987. as supplemented June 16, 1987 (TS 87-29)
Description ofamendment requester Tennessee Valley Authority (TVA.the licensee) proposed to modify the Sequoyah Nuc)ear Plant. Units 1 and 2 Technical Specifications (TS) to delete the reference to active motor-operated valves (MOVs) which willhave their thermal overload (TOL) protection devices bypassed. In addition, MOVs are also deleted which are no longer
, active, because of 10CPR Part 50, Appendix R considerations or because their movement is not required for safety. Several active MOVs which were previously omitted are added to the TS table. A typographical error ln the Unit 1 TS table is also corrected. The TS table is also reordered by system and valve number. By letter dated June 16, 1987.
TVAprovided four attachments which were Inadvertant)y omitted from the original submittal. This letter also provides additional justification for the proposed changes.
Basis forproposed no significant hazards consideration determinatianr The Commission has provided standards for determining whether a significant hazards determination exists as stated in10CFR 50.92(c). 10CFR 50.91 requires at the time a licensee requests an amendment. itmust provide to the Commitrslon its analyses, using the standards in Section 5082, about the issre ofno significant hazards consideration. Therefore, in accordance with 10 CFR 50.91 and 10 CPR SO.Ig, the licensee has performed and provided the followinganalysis.
(t)ls!he prababi%ty ofan acauncnce or the consequences ofan accident previous/y eva/uated /n the sofety ana/ysis report significant/y inoreased7 No. The Intent of Regulatory Guide {RG) 1.106, "Thermal Overload Protection for Electric Motors on Motor&pcratcd 9'elves," Is to ensure that the rnotorwpcratcd valve (MOV)performs Its intended function under accident conditions and that some protection against MOV degradation is provided during norinai operations. As stated In Regulatory Position 1.a, continuously bypassing the thermal overload {TOL)device Is an acceptable method ofcnsurhg that the MOVperforms its safety function. MOVdegradation can be detected through the use ofMoto~crated Valve Analysis and Testing System (MOVATS)and the preventive maintenance (PM) programs under development.
MOVATS,PMs, and ASME Section XItesting willensure that the probability ofMOV
'ailures {moto~pcrator or valve) Is not increased. The intent ofRG 1.108 Is satisficcL MOVATStcsthg is significantly more reliable as a tool for detecting MOV degradation then are TOLdeviccL The changes to the table ere made to reflect the active MOVs withTOL devices In force. As such, there Is a decrease in the probability of occuncnce or the consequences ofan acctdcnt previously evaluated in the safety analysis report because of the hcrcase In MOVreliability.
(2) /s the parr@lb//ityforan 'occident ofa new or different type than evaluated prev/aus/yin the safety analys/rr rcpor..
creatcd7 No. The continuous bypassing of the TOL devices and the MOVATStesting end PM programs are simply an alternative method formecthg tbe htent ofRG 1.1OL The possibility of MOV.failure. either motor-operator or valve, Is not creat'cd by removal of the TOL devices because of the known sensitivities of the TOL devices. Component operation willremain the same In terms of its Intended function. The dclctions and changes to the table are made to reflect the active MOVs withTOL devices In force. Therefore.
the changes do not create the possibility for an acddcnt of a new or different type than evaluated previously in the safety analysis rcport.
(3) /s the margin ofsafety significantly reduced7 No. The bypass of the TOLdevice is to ensure that the MOVs willp'crform their Intended function under accident conditions.
Also, testing Is performed to identify MOV dcgradritton, meeting the intent of RG l.lee:
and MOVATStesting Is more sensitive for detecting MOVdegradation than are the TOL
Federal Register / Vol. 62, No. 126 / Wednesday, July 1, 1987 / Notices 24561 devices. This results In higher MOV reliability.The changes to the table are made to reflect the active MOVs with TOLdevices in force. Thus. there is an Increase fn the margin of safety because of the higher MOV reliability.
The staff has reviewed the licensee's no significant hazards consideration determination and agrees with the licensee's analysis. In addition. the Commission hes provided certain examples (51 FR 7751) of amendments not likelyto involve significant hazards considerations. The remaining proposed changes to add active MOVs omitted from the TS table, to correct a typographical error in the Unit 1 TS table, and to reorder the TS table by system and valve number are encompassed by example (1), e.g., a purely administrative change to TS: for example. a change to achieve consistency through out the TS, correction of an error, or a change in nomenclature. Therefore, the staff proposes to determine that the application for amendments involves no significant hazards consideration.
Local Public Document Boom location: Chattanooga-Hamilton County Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
Attorneyforlicensee: General Counsel, Tennessee Valley Authority, 400 West Summit HillDrive, E11 B33, Knoxville,Tennessee 37902.
NRC Assistant Director: John A.
Zwolinski Tennessee Valley Authority. Docket Nos. SM27 and 5M28, Sequoyah Nuclear Plant, Unite 1 and 2, Hamilton County, Tennessee Date ofamendment requests: May 18.
1987. supplemented on June 4, 1987 (TS 87-34)
Description ofamendment requests:
Tennessee Valley Authority (TVA.the licensee) proposed to extensively modify Section 8, "Administrative Controls," of the Sequoyah Nuclear Plant. Units 1 and 2 Technical Specifications (TS). For clarity, the proposed change hes been divided into five areas as follows: (1) Office of Nuclear Power (ONP) Reorganization, (1) plant Operations Review Committee (PORC). (3) Independent Safety Evaluation Group (ISEG) ~ (4)
Radiological Assessinent Review Committee (RARC), and (5) Fire Brigade Members. A separ'ate description for each area ls provided below.
(t) ONP Reorganization
>iiministrative change is being made to change the title of plant Superintendent to Plant Manager. The Plant Manager
>'auld report directly to the Site IIirector. The title of Assistant Director of Nuclear Power (Operations) ls being changed to Site Director. The chief officer of the ONP ls now titled Manager ofNuclear Power and,the Health Physicist ls now titled Site Radiological Control Superintendent, The reference to the offsite organization for radiological environmental monito'ring program and dose calculations is being removed. The site organization includes the new Plant Operation Review Staff (PORS). PORS is a staff reporting to the
~ Assistant Plant Manager.
(2) PORC The proposed changes affecting PORC would delete several items from PORC responsibility and place them under the new "technical review and control" process. The process would establish required "independent qualified review" and cross-disciplinary review and approval to support changes currently under PORC review responsibility. PORC would then be responsible for providing an oversight review of selected safety evaluations reviewed under the new process. This change would also clarify PORC responsibilities forreview of violations of the TS and reportable events. By letter dated June 4, 1987, TVA
'ithdrew a portion of the May 18, 1987 submittal retaining the requirement for proposed TS changes being reviewed by the PORC. Also withdrawn, as stated in the June 4, 1987 letter, is the proposed removal ofSpecial tests and experiments from PORC review.
Therefore. the June 4. 1987 letter narrows the scope of the original request, The May 18, 1987 submittal also proposes changes to accurately reflect the titles ofthe regular PORC members.
Temporary procedure changes are now provided for in only unusual circumstances, provided final approval is implemented within 14 days. A typographical error ls also being corrected.
(3) ISEG The proposed composition of ISEG for the Sequoyah Nuclear Plant (SQN) is a change from five dedicated full-time onsite engineers to three dedicated full-time onsite engineers supplemented by two full-time engineers located in corporate headquarters and sh'ared by all TVAnuclear facilities.
ISEG would report to the Director of Nuclear Safety and Licensing Division.
(4) RARC-An administrative change is proposed to change the titles of the members of RARC to be consistent with the positions responsible for those functions under the new organization.
~ RARC would provide reports to the
'Manager ofRadiological Control (formerly the Chief, Radiological Hygiene Branch) and the Plant Manager.
(5) Fire Brldgade Members For the Sequoyah Nuclear Plant Unit 1 TS only, TVAproposed to change the specificatlon of the shift supervisor as one of the three excluded personnel from membership in the fire brigade.
Unit 2 has previously been changed to this requirement.
Basis forproposed na significant hazards consideration 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 about the issue ofno significant hazards consideration, Therefore, ln accordance with 10 CFR 50,91 and 10 CFR 50.92, the licensee has performed and provided the followinganalysis.
(tlIs the probabflity ofan occurrence or the consequences ofan occident previously evaluated in the safety analysis report significantly increased7 No. This change Is intended to accurately reflect the changes to the Administrative Controls Section (section
- 8) of the technical specifications due to the realignment of the ONP. The functions spectfied in section 8 important to the safe operation ofSQN have not been altered or delete*The changes to this section merely reflect the new positions that hold the expertise to perform these functions. The rol&ofPORC, RARC. NSRB, and the ISEG are advisory roles to provide technical assistance to those individuals changed with the responsibility of safe operation of this facility.'Ibischange better Identifies those positions occupied by Individuals qualified io provide this technical assistance. There are no hardware. procedure, personnel, or analysis changes represented by thi~
amendmont proposal which adversely affect the probability of occunence or the consequences of an accident previously evaluated In the safety analysis report.
The proposed changes to PORC responsibilities willsignificantiy reduce the number of Items that must be reviewed by PORC, thus allowing the PORC members to focus their attention on the more important safety-related issues. Those Items being removed from PORC review willstill receive a detailed technical review by qualified individuals.
The revision to unit one fire brigade.
membership serves only to make the technical spectfications for both units consistent. This change, In excluding the shift supervisor from membership in the brigade.
ensures that the Individual specified In e.t.s as being responsible for the control room command function Is not distracted by this peripheral reeponsibflity. This change does not adversely atlect the safety analysis.
(2) Is the possibilityforon accident of o new or different type than evaluated previouslyin the safety analysis report cteated7No. This change represents. only a change in the adininistrative process In that new positions and responsibilities are identlfied for review functions of
Federal Regbter / Vol. 52, No. 126 / Wednesday, July 1, 1M7 / Notices organizations such e's PORC, RARC. NSRB, end ISEG. This review functhn willcontinue to be performed by these individueh who are technically competent to perform these review or advisory roles; therefore, the potential for the hcreese pf e possibility of an accident or e new or diif<<reit type accident from an hied equate review Is reduced rather than increased due to having the appropriate personnel designated for these functions.
(3) Ie the margin ofsafety significantly educed? No. The changes tn this amendment proposal serve only to clarify those positions responsible for key safety functions specified in section 6 of technical spectficsttonL No function has been Impaired or deleted by this change. On the contrary, these functions are enhanced by additional clarity afforded by the reorganization.
The staff has reviewed the licensee's no significant hazards consideration determination and agrees with the licensee's analysis. Therefore, the staff proposes to determine that the application for amendments involves no significant hazards consideration.
Loca/Public Document Boom location: Chattanooga-Hamilton County Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
Attorneyforlicensee: General Counsel, Tennessee Valley Authority, 400 West Summit HillDrive. E11 B33.
Knoxville, Tennessee 37902.
NBCAssistant Director: John A.
Zwolinskl Vermont Yankee Nudear Power Corporation, Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, Vermont Dateof application foramendment:
August 5, 1983 (48 FR 49596) and March 4, 1985 Description ofamendment request.
The licensee has requested that a requirement for reactor protective system RPS power protection panel operability be added to the Technical SpecIGcations. and that a requirement be added to the Technical SpecIGcations to trip an RPS protective panel ifit is found inoperable. 'I1ie licensee has also proposed surveillance teste ofRPS protective panel overvoltage, undervoltage, and underfrequency relays be added to the Technical Specifications. The licensee has proposed these changes in order to meet NRC requirements for Technical Specification assurances of'operability for recently-Installed RPS power protection panel operability. The RPS power protection panels were installed to alleviate NRC concerns that voltage could be varied sufGcieatly by a seismic'vent to cause failure ofthe RPS.
Bosis forproposed no significant hazards consideration determination:
The Commission has provided guidance concerning the application of the standards in 10 CFR 50.92 by providing certain examples (51 FR 7751). One of the examples (ii)ofactions not likelyto Involve a significant hazards consideration is a change which constitutes an additional limitation, restriction. or control not presently included in the Technical Specifications, for example, a more stringent surveillance requirement.
. 'he changes included in this
'pplication add limitations not presently included in the Technical Specifications.
The proposed changes add words to the Technical Specifications requiring that surveillance be performed on RPS power protection equipment and action be taken ifequipment fs found inoperable.
Therefore these changes are similar to example (ii).
Accordingly, the Commission proposes to determine that the proposed amendment does not involve a significant hazards consideration.
Local Public Document Boom Location: Brooks Memorial Library. 224 Main Street, Brattleboro, Vermont 05301.
Attorneyforlicensee: John A.
Ritscher, Esq., Ropes 8 Gray. 225 Franklin Street. Boston, Massachusetts 02110.
NBCPleat Director: VictorNerses, Acting Director NOTICE OF ISSUANCE OF AMENDMENTTO FACILITY OPERATING LICENSE During the period since publication of the last bi-weekly 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 ofthe AtomicEnergy Act of 1954. as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate Gndings 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 ofConsideration ofIssuance of Amendment to Facility Operating License and Proposed No Significant Hazards Consideratioa Determination and Opportunity for Hearing in connection with these actions was published in the Fedora Reghiter as
, indicated. No request for a hearing or petition for leave to Intervene was Gled 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 5122(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, 1717 H 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 Regulatory Commission, Washington. DC 20555. Attention:
Director, Division of Licensing.
Cleveland Electric Illuminating-Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania Power Company. Toledo Edison Company, Docket No. 5M40, Perry Nuclear Power Plant, Unit No. 1, Lake County, Ohio Date ofapplication foramendment:
December 15, 1986, as amended February 10, 1987 Briefdescription ofamendment: The amendment makes several editorial changes and corrections. and administrative changes, including deleting the requirement that NRC approval ofitems involvingunreviewed safety questions shall be obtained prior to licensee internal approval for implementation. The amendment also makes several technical changes: using ofUnit 2 divisional batteries as an alternate DC power source forUnit 1 shutdown. increasing the number and changing the location of drywell average air temperature instruments, changing the automatic depressurization system instrument air low pressure alarm setpoint. deleting an obsolete footnote, and changing the containment vacuum breaker isolation valve opening setpoint The Commission has denied the portion of the amendment application requesting an administrative change related to corporate aad staff organization charts. ANotice of Denial ofAmendment has been published separately in the Federal Register.
Date ofissuance: June 9. 1987 Effective date: June 9, 1987
~ Amendment No. 6
Federal Reghter / Vo 5?
No. 128 / Wednesday, July 1, 19S7 / No'Ucas Fac/I(ty Operating License No. NPF-SS. This amendment revised the Technical SpecificaUons, Dote ofinitialnollce In Federal Register. March 12, 1987 (52 FR 7878)
The Commission's'ge]ated evaluation of the amendment is cotttained in a Safety Evaluation dated June 9, 1987.
No significant hazards consideration comments received: No, Local Public Document Room locotioru Perry Pubfic Library, 37S3 Main Street, Perry, Ohio 44081 Duke Power Company, Docket Noe. 50-289, 50-270 and 50-287, Oconee Nuclear Station. Units 1, 2, and 3, Oconee County, South Carolina Dote ofapplication foramendments:
February 10, 1986, as supplemented on August 20, 1986 Briefdescription ofamendments:
These amendments revise the Station's common Technical Specifica Uons (TSs) to (1) revise TS 3.1.12.1(a), (b) and (d) to indicate that three subcooling margin monitors are now available over the previous two monitors and refiect the actual plant design and (2) delete TS 3.1.12.1(c) to no longer require a 3&day report for outages of less than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of the Operational AidComputer. Also TS 3.1.12.1(d] has been redesignated as TS 3.1.12.1(c) because TS 3.1.12.1(c) has been deleted.
Date ofIssuance:,
June 8, 1987 Effective date: June 8, 1987 Amendmeht Nos.: 159, 159 and 156
, Facility Operating Licenses Nos.
DPR4S, DPRA7, and DPRSS.
Amendments revised the Technical Specifiications.
Date ofinitialnoticein Federal Register. May 6, 1961 (52 FR 16942)
The Commission's related evaluation of the amendments is contained in e Safety Evaluation dated June 8. 1987.
No significant hazards consideration comments received: No Local Public Document Room Iocatioru Ocone'e County Library. 501 West Southbroad Street, Walhalla, South Carolina 29691 Florida Power and LIght Company, et al Docket No. SM89, St. Lucie Hant, Unit No. 2, St. Lucie County, Horida Dote ofapplication ofamendment:
April21. 1986, as supplemented October
- 7. 1988 and April10, 1987.
Briefdescription ofamendment: The '
amendment deletes License Condition
. 2.C.(tS] which limited the burnup of spent fuel in the spent fuel pool to 38,000 MWD/MTU.
Dale ofIssuance: @jay 29. 1S87 Effective Bate: May 29, 1967 Amendment No, 21 i
FacllllyOperaffng License ¹. NPF-18; Amendment revised the License.
Date ofInitialrialice in Federal Register. May 21, 1986 (52 FR 18882)
The licensee provided addiUonal informaUon subsequent to the iniUal noUce published in the Federal Reghiter.
The October /, 1988 submittal provided ciarificaUon. The April10, 1988 submittal provided minor conecUons to the analysis. This additional informaUon does not alter our proposal ofno significant hazards consideration.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 29. 1987.
No significant hazards consideration comments received: No.
Local Public Document Boom location: Indian River Junior College I1brary. 3~~
Vtrgina Avenue, Ft. Pierce, Horida.
Florida Power and Ligb Company, Docket Nos. 50-250 and 50-251, Turkey Point Plant Units 3 and 4, Dade Countye Horida Date ofapplication foramendments:
May 7, 1988, as supplemented on February 20, 1987, and April23, 1987.
Briefdescription ofamendments:
These amendments delete the specifications for the Auxiliary Feedwater (AFW) System and the Condensate Storage Tanks (CST) in current Technical SpecificaUon 38, Steam and Power Conversion Systems.
Requirements for the AFW System and CST are included in the new Technical Specifications 3,18 and 3.19. The specifications provide explicit limiting conditions for operation (LCO),
applicability requirements, and Action requirements for operation of the AFW System and CST. The format (i.e LCO.
appiicabfiity, Action requirements) is that ofNUREG4452. Standard Teel&ca l Specifications. for Westinghouse Pressurized Water Reactors (WSTS), although the requirements in the proposed SpecificaUons differ from the WSTS because of the uniqueness of the Turkey Point Plant AFW System design (i.e.,
shared system, three turbine driven pumps, etc.)
The amendments also provide surveillance requirements for the CST which were not included in the existing Technical SpectGcaUons, correct errors in the valve numbers for two primary coolant system pressure isolation valves. and update the Bases to support the changes for the AFW System and CST.
Date ofissuance: June 8, 19&7 Effective date. June 8, 1981 Amendment Nos, 124 and 118 Faci/lty Operating Licenses Nos.
DPR41 and DPR41; Amendments revised the Technical Specifications, Date ofInitialnotice in Federal Reciter. June 18, 1986 (61 FR 22236) and
, renoUced on May 8, 1961 (62 PR 16946)
" The Commission's related evaluation of the amendments Is contained in a Safety Evaluation dated June 8, 1987.
No slgnificant hazards consideration comments received: No LocalPublic Document Room location: Environmental and Urban AffairsLibrary, Florida International University, Miami, Florida 331S9.
GPU Nuclear Corporation, et aL, Docket No. 50-289, Three Mle Ishnd Nuclear Station, Unit No 1, Dauphin County, Pennsylvania Dote ofapplication foramendment:
February 3, 1987 Briefdescription ofamendment; This amendment revised the Technical Specifications to account for a new condenser vent stack iodine sampler.
The new iodine sampler provides for continuous sampling unrestricted areas to determine compliance with 10 CFR 20 and 10 CFR 50. Appendix L The amendment also made an editorial change in Table 4.22-2 Dote ofissuance: June 8, 1987-Effective date: June 8, 198/
Amendment No '130 Facility Operating License ¹. DPR-
- 50. Amendment revised the Technical Specifications.
Date ofinitialnodcein Federal Regbiter. April22, 1987 (52 FR 13337)
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 8, 1987.
No significant hazards consideration comments received: No.
LocalPublic Document Room Iocationi Government Publications Section, State Library of Pennsylvania, Education Building, Commonwealth and Walnut Streets, Harrisburg, Pennsylvania 17126 Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 5M'nd 50488, Edwin I. Hatch Nuclear Plant, Units 1 and 2. Appling County, Georgia Date ofapplication foramendments:
June 20, 1988 as supplemented July 22, 1986 and January 2, 1S87 Briefdescription ofamendmeats: The amendments modify the Technical Specifications to permit operation with only one recirculaUon loop in operation and to bn'plement the jet pump surveillance recommendations of
24564
- ~
.... ~
Federal Register / Vol. 52, No. 128 / Wednesday, J jy 1, 1987 / Notices NUREG/CR-3052 "BWR Jet Pump-Assembly Failure."
Date ofissuance: June 10, 1987 Effective date: June 10, 1987 Amendment ¹si141 and 77 Facility Operating License Nos. DPB-57 and NPF-$. Amendments revised the Technical Specifications.
Date ofinitialnotice in Federal
,Register. August 27, 1986 (51 FR 30572}
The Applicant's January 2. 1987.
supplement was merely to correct a typographical error.
,The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 10, 1987 No significant hazards consideration comments received: No Local Public Document Room lacationi Appling County Public Library.
301 City Hall Drive, Bax!ey. Georgia 31513 Gulf States Utilities Company, Docket No. 50458, River Bend Station, Unit 1 West Peliclsne Parish, Louisiana Date ofapplication foramendmenti january 28. 1987 as supplemented February 13 and April16. 1987.
Biiefdescription ofamendment: The amendment modifjed the Scram Discharge Volume Water Level - High trip setpoint for the Float Switches LSN013A, B. C and D in the Technical Specifications.
Date ofissuance:
June 8, 1987.
Fffective date: June B, 1987.
Amendment No. 6 Facility Operating License No. NPF-
- 47. This amendment revised the Technical Specifications.
Dale ofinitialnotice in Federal Register. May 8, 1987 (52 FR 16948).
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 6, 1987.
No significant hazards consideration romments received: No Local Public Document Room location: Government Documents Department. Louisiana State University.
Baton Rouge. Louisiana 70803 indiana and Michigan Electric Company, Docket Nos. 5M15 and 5M1B, Donald C. Cook Nuclear Plant, UnIt Nos. 1 and 2, Bemen County, Michigan Date ofapplication foramendmentsi t larch 26, 1987 (Partial).
Briefdescription'of amendmentsi The amendments changed the Technical Specifications (TS) to revise the refueling operation boron concentrations. increase the boron concentrations in the refueling water storage tank (RWST) and accumulators, increase the usable water volumes required in the RWST, increase the RWST temperature at all times to prevent precipitation, make the moderator temperature coefficient a ramp function withpower rather than a step function, and add footnotes such that addition of water from the RWST does not constitute a boron deletion.
The remaining items from the licensee's March 26, 1987 submittal willbe the subject of a separate action.
Dote ofissuance: June 10. 1987.
Effective date: June 10, 1987.
Amendment Nos,'111 and 94, Foci%'ty Operating License Nos. DPB-58 and DPR-74. Amendments revised the Technical Specifications.
Date ofinitialnotice in Federal Register: May 8, 1987 (52 FR 18949).
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 10, 1987.
No signiTicant hazards consideration comments received: No.
Local Public Document Room location: Maude Preston Palenske Memorial Library, 500 Market Street. St.
Joseph. Michigan 490SS Louisiana Power and Light Company, Docket No. 50482, Waterford Steam Electric Station, Unit 3, St. Charles Parish, Louisiana Date ofamendment requesti February 23, 1987 Briefdescription ofamendment: The amendment revised the Technical Specifications by: (1) raising the emergency feedwater initiation aetpoint from 30%,to 36.3% ofwide range level; (2) raising the required refueling water storage pool level from 82% to 83%; and (3) raising the required safety injection tank level from 80% to 81%.
Dote ofissuance: June 1S, 19S7 Effective date: June 15, 1987 Amendment No,'9 FacilityOperating License No. NPF-J8. Amendment revised the Technical Specifications.
Date ofinitialnotice in Federal Register. March 25, 1987 (52 FR 9574<}
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 15, 1987.
No significant hazards consideration comments received: No.
Local Public Document Room location." University ofNew Orleans Library. Louisiana Collection. Lakefront.
New Orleans, Louisiana 70122 Maine Yankee Atomic Power Company,
'Docket No.5M09, Maine Yankee Atomic Power Station, Lincoln County, Maine Date ofapplication foramendment:
February 24. 1987 Briefdescription ofamendment: This amendment modifies the Technical Speciifiications to refiect revised L'oss of Coolant (LOCA)Limits.These revised LOCAUmlts were determined using a revised delta P injection penalty factor during the reflood phase ofLOCA's and limitingaxial power shapes. The revised
~ delta P injection penalty factor and the
'inethod to select%a limitingaxial power shapes were proposed and justified in the Maine Yankee Atomic Power Company (MYAPCo) letter to NRC dated November 10, 198L Theaq proposed revisions to the Emergency
'ore Cooling system (ECCS) Evaluation Model were reviewed and found acceptable in the NRC letter to
, MYAPCo dated January B. 1987.
Date ofissuance: June 15. 1987 Effective date: June 15, 1987 Amendment No.! 98 Facility Operoting License No. DPR-
- 88. Amendment revised the Technical Specifications.
Date ofinitialnoticein Federal Register: March 25, 1987 (52 FR 9577)
The Commission's related evaluation of the amendment is contained'in a Safety Evaluation dated June 15, 1987.
No significant hazards consideration comments received: No Local Public Document Boom location: Wiscasset Public Library. High Street Wiscasset. Maine 04578 Northeast Nuclear Energy Company, et al., Docket No. 50-245, Millstone Nuclear Power Station, Unit No. 1, New London County, Connecticut Date ofapplicatian foramendment:
January 20. 1987 Briefdescription ofamendment: The amendment adds six (8) fire protection water suppression systems to the list of fire protection systems in Technical Specification 3.12.B.1. Pour (4) of the proposed suppression systems are located in the turbine building. The other two (2) are located in the reactor building. The proposed additions are new requirements for suppressing fires in the lube oil systems of the condensate booster pumps. reactor feedwater pumps and motor generator sets, as well as in the curbed area of the motor.
generator sets, turbine building unloading area and the mezzanine level cable tray,
'ate ofissuance: June 5, 1987 Effective date: June 5, 1987 Amendment No, 3 FacilityOperaliiig License No. DPR-2I. This amendment revised the Technical Specifications.
Date ofinitialnoticein Federal Register. March 12. 1987 (52 FR 7686).
The Commission's related evaluation of the amendment is contained in a
. Safety Evaluation dated June 5, 1987.
Federal Register / Vol. 52, No. 128 / Wednesday, July 1, 1987 Notices No significant hazards consideration comments received: No.
Local Public Document Boom lacatfam Waterford Public Library, 49 Rope Feny Road, Waterford, Connecticut 08385.
Northeast Nuclear Energy Company, et alDocket No. 50423, Millstone Nuclear Power Station, Unit No, 3, Town of Waterford, Connecticut Date ofapplication foramendment:
December 10, 1986 Briefdescription ofamendment: The amendment revised the Technical Specifications Sections 4.6.1.2.d, 4.8.1.2.g. 4.6.1.1.2 and Technical Specification Bases Section 3/4.8.1.7 to delete the requirement to leak test the containment purge supply and exhaust isolation valves every six months.
instead, these valves would be leak tested at intervals no greater than 24 months in acordance withTechnical Specification Section 4.6.1.2.d and 10 CFR 50. Appendix J in conjunction with a valve seat replacement program.
Date ofissuance: June 1S. 1987 Effective date: June 15, 1987 Amendment No, 5 Foci%'ty Operating License Na. DPR-R/. Amendment revised the Technical Specifications.
Date ofinitialnoticein Federal Register. March 12, 1981 (52 FR l688)
The Commission's related evaluation uf the amendment is contained in a Safety Evaluation dated June 15, 1981.
No significant hazards consideration comments received: No.
Local Public Document Room location: Waterford Public Library, 49 Rope Ferry Road, Waterford.
Connecticut 08385 Northeast Nudear Energy Company, Docket No. 50-423, Millstone Nuclear Power Station, Unit No. S, Tawn of Waterford, Connecticut Date ofamendment requeati December 5, 1988 Description ofamendment request:
The amendment revised Millstone Unit No. 3 Technical Specification Figure 62-2, Unit Organization, by changing the depiction of the Millstone Station Services Organization as individual Security, Quality Services and Radiological Services groups reporting to the Station Services Superintendent to depiction as one group of staff including QA, Security, Health Physics, Chemistry and other Site Support Services Staff reporting to the Station Services Superintendent.
Dote ofisauancei June 1S, 1981 Effective date: June 15, 1987 Amendment No,. 6 Facility Operating License No, NPF-4R Amendment revised the Technical Specifications.
Date ofinitialnotice In Federal Registen March 12, 1987 (S2 FR 7688).
The Commission's related evaluation
'f the amendment is contained in a Safety Evaluation dated June 15, 198'l.
No significant hazards consideration comments received: No, Local Public Document Boom locatiom Waterford Public Library, 49 Rope Perry Road, Waterford, Connecticut 06385 Pacific Gas and Electric Company, Docket No. SMZS, Diablo Canyon Nuclear Power Plant, San Luis Obispo County, Califonda Date ofapplication foramendmen4 March 17, 1987, as supplemented May 6, 1981.
Briefdescription ofamendment: The amendment extends the time for submittal ofa steam generator tube rupture analysis to April198L Effective date: June 12, 1987 Amendment Na,'12 Facility Operatiiig License Na DPR-8?: Amendment revised the license.
Date ofinitialnotice in Federal Register. May 1? 1987 (51 FR 17884)
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 12, 1987.
No significant hazards consideration comments received: No, Local Public Document Boom location: California Polytechdc State University Library, Government Documents and Maps Department, San Luis Obispo, California 93401, Pacific Gas and Electric Company, Docket Nos. 50-275 and 5M2$, DiaMo Canyon Nuclear Power Plant, Unit Nos.
1 and 2, San Luis Obispo County, Califonda'ate'af applicatfan foramendments:
February 10, 1987 Briefdescription ofamendments:
These amendments allow replacement of fuel rods in fuel assemblies with flier
~ rods or vacancies on a limited basis provided that such replacement is demonstrated to be acceptable by a cycle-specific reload analysis.
Date ofissuance: June 8,.1981 Effective date: June 8, 1987 Amendment Naa: 13 and 11 Facility Operating Licenses Naa.
DPR~ and DPR~ Amendments revised the Technical Specifications.
Date ofinitialnoticein Fedmal Register. April2Z, 1987 (52 FR 13344).
The Comndssian's related evaluation
~ ofthe amendments is contained in a Safety Evaluation dated June 8. 1987 No siydficant haxards consideration comments received: No LocalPublic Document Boom locatlom Callfonda Polytechnic State University Library, Govenunent
,: Documents and Maps Department, San Luis Obispo, Califonda 93401 Paclfic Gas and Electric Company, Docket Nos. 50.275 and 5M', IMablo Canyon Nudear Power Plant, Unit Nos.
1 and? San Luis Obispo County, '.
Califonda Date ofapplication foramendments:
March 25, 1987. as supplemented May 28, 1987 Briefdeacriptfon ofamendments: The amendments revised the Technical Specifications to accommodate Cycle 2 and later operation ofUnit 2, and Cycle 3 and later operation of Unit 1.
Date oflaauancei June 12, 1987 Effective date: June 12, 1987 Amendment Noa: 14 and 13 Facility Operating Licenses Noa.
DPR~ and DPR~ Amendments revised the Technical Specifications.
Date ofinitialnotice in Federal Register. May 6, 1987 (SZ FR 16949)
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 1Z. 1987 No significant hazards consideration comments received: No L'acal Public Document Room location: California Polytechnic State University Library, Govenunent Documents and Maps Department, San Luis Obispo, Califonda 93401 Pennsylvania Power and Light Company, Docket No. 5MN, Susquehanna Steam Electric Staten, Unit? Luxerne County, Pennsylvaida Date ofapplication foramendment:
March 21, 1988, as revised April18,'1988.;
March 2, and April3, 1987 Briefdescriptio ofamendment This amendment revised the SSES Unit 2 Techdcal Specifications to include operational control on equipment which must be operable to ensure proper functioning ofthe newly installed drywell cooling fans.
Date offaauancei June 5, 1987 Effective date: June 5, 1981 Amendment No. 38 Facilfty Operating License No, NPF-2? This amendmant revised the Technical SpecificationL Date ofinitialnotice fn Federal Reghter. May 7, 1988 (51 FR 18932)
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 5. 1987.
No sfgnificant hazards consideratiea comments received: No
2456S Federal Register / Vol. 52, No. 126 / Wednesday, July 1, 1987 / N4tices Local Public Document Room location: Osterhout Free Library, Reference Department, 71 South Franklin Street, Wilkes.Barre, Pennsylvania 18701.
'ower Authority oF'@e State"of New York, Docket No. 50-288, Indian Point Unit No. 3, Westchester County, New York Dote ofapplication foramendment:
March 10. 1987.
Briefdescription ofamendment: The amendment revises the Technical Specification requirement for control bank insertion limits. The revision is being made to reflect a more conservative insertion position for the C and D control banks.
Date ofissuance: June 8, 1987
~
Effective dale: June 8, 1987 Amendment Na: 75 Facilities Operatic License No: DPR-
&t:Amendment revised the Technical Specifications.
Dote ofinitialnotice in Federal Register. May 8. 1987 (52 FR 16954)
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 8. 1987.
No significant hazards consideration comments received: No Local Public Document Room location: White Plains Public Library, 100 Martine Avenue, White Plains. New York, 10610, Public Service Electric gi Gas Company.
Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey Date ofopplication foromendment:
April3. 1987. as supplemented May 8.
1987 Briefdescription ofamendment: The amendment has granted relief from an ASME Boiler and Pressure Vessel Code, Section XLvalve leakage test requirement for the Hope Creek Generating Station. The amendment also extended the current Technical Specification surveillance intervals for 27 reactor coolant system pressure isolation valves and primary containment isolation valves, on a one-time-only basis. from once every 18 months and once every 24 months, until the first refueling outage (currently scheduled to begin on February 1. 1988).
The code relief allows leak tests of the
'7 valves. required by the code to be performed no less than once every two years. to be deferred until the first refueling outage.
Date ofissuance:.June
- 9. 1987 Effective date: June 9. 1987 Amendment Na 4 Facility Operatic License No. NPF-
- 57. This amendment revised the Technical Specifications.
Date ofinilialnoticein Federal Register. (52 FR 16954) May 6, 1987 The Commission's related evaluation of the amendment is contained in a Safety, Evaluation dated June 9. 1987.
No significant hazards consideration comments received: No LocalPublic Document Room location: Pennsville Public Library. 190 S. Broadway. Pennsville, New Jersey 08070 Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, Callaway County, Missouri Date ofopplication foramendmenl:
March 27, 1987.
Briefdescription ofamendment: The amendment modified Section 5.3.1 of the Technical Specifications to allow for limited replacement offuel rods with fillerrods or vacancies ifsupported by a cycle-speciTic reload analysis, Date ofissuance: June 6, 1987.
Effective date: June 8. 1987.
Amendment No,'4.
Facility Operating License No. NPF-
- 30. Amendment revised the Technical SpeciTications.
Date ofinitialnoticein Federal Register. April22, 1987 (52 FR 13350)
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 8. 1987.
No significant hazards consideration comments received: No.
Local Public Document Room location: Callaway County Public Library, 710 Court Street, Fulton, Missouri 65251 and the John M. Olin Library, Washington University, Skinker and Lindell Boulevards, St. Louis, Missouri 63130.
NOTICE OF ISSUANCE OF AMENDMENTTO FACILITY OPERATING LICENSE ANDFINAL DETERMINATIONOF NO SIGNIFICANTHAZARDS CONSIDERATIONAND OPPORTUNE FOR HEARING (EXIGENTOR EMERGENCY CIRCUMSTANCES)
During the period since publication of the last bi-weekly notice. the Commission has issued the following
, amendments. The Commission has determined for each of these amendments that the application for the amendment complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations.
The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter L which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the date the amendment was needed, there was
, not time for the Commission to publish,
~for public comment before issuance. its usual 30-day Notice ofConsideration of, Issuance ofAmendment and Proposed No Significant Hazards Consideration Determination and Opportunity for Hearing. For exigent circumstances.
the Commission has either issued a Federal Register notice providing opportunity for public comment or has used local media to provide notice to the public in the.
area surrounding a licensee's facilityof the licensee's application and'of the Commission's proposed determination of no significant hazards consideration.
The Commission has provided a reasonable opportunity for the public to comment, using its best efforts to make available to the public means of communication for the public to respond quickly, and in the case of telephone comments, the comments have been recorded or transcribed as appropriate and the licensee has been informed of the public comments.
In circumstances where failure to act in a timely way would have resulted. for example. in derating or shutdown of a nuclear power plant or in prevention of either resumption ofoperation or of increase in power output up to the plant's licensed power level. the Commission may not have had an opportunity to provide forpublic comment on its no significant hazards determination. In such case. the license amendment has been issued without opportunity for comment. Ifthere has been some time forpublic comment but less than 30 days, the Commission may provide an opportunity for public comment. Ifcomments have been requested. it is so stated. In either event, the State has been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an amendment immediately effective, notwithstanding the pendency before it of a request for a hearing from any person. in advance of the holding and completion of any required hearing, where it has determined that no significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has made a final determination that the amendment involves no significant hazards consideration. The basis for this determination is contained in the documents related to this action.
Accordingly. the amendments have been issued and made effective as indicated.
~
~
Federal Register / Vol. 52, No. 128 / Wednesday, July 1, 1987 / Notices 2466'1 Unless otherwise Indicated, the Commission has determined that these amendments satisfy the criteria for cutegorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), np environmental impact statement or environmental assessment need be prepared for these amendments. Ifthe Commission has prepared an environmental assessment ur der 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 artion see (1) the application for amendment, (2) the amendment to Facility Operating License, and (3) the Commission's related letter, Safety Evaiua!ion and/or Environmental Assessment.
as indicated. AH of these items are available for public inspection at th'e Commission's Public Document Room, 1717 H Street, NW. Washington, DC, and at the local public document room for the particular facilityinvolved.
A copy of items (2) and (3) may be obtained upon request addressed to the U.S. Nuc!ear Regulatory Commission.
Washington. DC 20555, Attention:
Director, Division of Licensing.
The Commission is also offering an opportunity for a hearing with respect to the issuance of the amendments. By July
- 31. 1987, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written petition for leave to intervene. Requests for a hearing and petitions for leave to intervene shall be filed in accordance with the Commission's "Rules of Practice for Domestic LIcensing Proceedings" in 10 CFR Part 2. Ifa request for a hearing or petition for leave to intervene is filed by the above date. the Commission or an Atomic Safety and Licensing Board, designated by the Commission or by the Chairman of the Atomic Safety and Licensing Board PaneL willrule on the request and/or petition and the Secretary or the designated Atomic Safety and Licensing Board wiH issue a notice of hearing or an appropriate
- order, As required by 10 CFR 2.714. a petition for leave to intervene shall set'orth with particularity the interest of the petitioner in the proceeding and how that interest may be affected by the results of the proceeding. The petition should specificaHY explain the reasons why intervention should be permitted
.with particder reference to the followingfactors: (1) the nature ofthe petitioner's right under the Act to be 3737 and the followingmessage made a party to the proceeding; (2) the addressed to (Project Directorft nature and extent of the petitioner's petitioner's name and telephone property, financial, or other interest in number; date petition was mailed; plant the proceeding, and (3) the possible name; and publication date and page effect of any order which may be
'iimber of this Federal Register notice.
entered in the proceeding on the
'A copy of the petition should also be petitioner's interest. The petition should sent to the Office of the General also identify the specific aspect(s) of the Counsel-Bethesda, U.S. Nuclear subject matter of the proceeding as to Regulatory Commission, Washington, which petitioner wishes to intervene, DC 20555, and to the attorney for the Any person who has filed a Petition for
- Hcensee, leave to intervene or who has been Nontfmely filings of petitions for leave admitted as a Party may amend the to intervene, amended petitions, Peti<<on without requesting leave of the supplemental Petitions and/or requests Board uP to fifteen (15) days Prior to the for hearing wIH not be entertained first Prehearing conference scheduled in absent a determination by the the Proceedin8, but such an amended Commission the presiding oflicer or the Petition must satisfy the sPecificity Atomic Safety and Licensing BoarrL that requirements described above.
Not later than fifteen (15) days prior to the Petition and/or request should be the first prehearing conference granted based upon a balancing of the scheduled in the proceedin8 a petitioner factors specified in 10 CFR 2.714(a)(1)(i)-
shall file a supplement to the petition to intervene which must include a list of The Cleveland Electric IHuminating the contentions which are sought to be Company Duquesne LIght Company, litigated in the matter, and the bases for Ohio Edison Company Pennsylvania each contention set forth with Power Company, Toledo Edison reasonable speciifiiCit. Contentions shall Company Docket No 5+440 perry be limited to matters within the scope of Nuclear P~~e~ Plant Unit No the amendment under consideration. A C ~tOhio petitioner who fails to file such a supplement which satisfies these
'ate ofapplication foramendment:
requirements with respect to at least one'ay 29, 1987 contention wiH not be permitted to Briefdescription ofamendment: The participate as a party.
amendment revises the steam tunnel Those permI<<ed to intervene become and twbine building main steam line arties to the proceeding, subject to anY high temperature trip setpoints and imitations in the order granting leave to afiowable values, items 2.f, 2 g. 2 h 4 f.
intervene, and have the opportunity to 4.g, 5.f and 5.g ofTable 3.3.2-2 of the par<<clpate fullyin the conduct of the Technical Speclficatlons.
hearing. Including the opportunit Date ofissuance: June 10, 1987 present evidence and crosiHrxamlne witnesses.
Ejective date: May 29, 1987 Since the Commission has made a Amendment Na. '7 final determination that the amendment FacilityOperating License No, NPF-involves no significant hazards, 5tr. This amendment revised the consideration, ifa hearing is requested, Technical Specifications.
it wiHnot stay the effectiveness of the Public comments requested as to amendment. Any hearing held would proposed no significant hazards take place while the amendment is in cansideratiant No. The Commission's effect.
related evaluation of the amendment, A request for a hearing or a Petition consultation with the State ofOhio, forleave to intervene must be filed with finding of emergency clrcunistances, and Secretary of fi e Co~ssionl U.S'inal determination ofno sIgnificant Nuclear Regulatory Commission, hazards consideration are contained In Washington, DC 20555, Attention:
a Safety Evaluation dated May 29, 1987.
Docketing and Service Branch. or may
.be defivered to the Commission's pubfic Attorneyforlicensee: Jay Silb rg.
Docirment Room, 1717 H Street, NW, Esq Shaw, PIttman Patte 8i Washington, DC by the above date,
~wb6 ge<~ N Street. NW, Where petitions are filed during the last Washington DC N037 ten (10) days of the notice period, lt Is requested that tho petitioner promptly so location: Perry Pubfic LIbrary 3753 Main inform the Commission by a toll-free Street Perry Ohio 44081.
telephone call to Western Union at (800)
NBCProject Director; Martin J.
3254000 (in Missouri (800) 3424700).
VIrgQIo.Acting.
The Western Union operator should be, Dated at Bethesda, Maryland ibis zsth day given Datagram ldentification Number offune 1087. '
b 1
July 13, 1987
'DI'STRI'BUTION Docket File PDR LPDR PD5 Rdg JLee (2)
DOCKET NO(S).
50-275/323 Hr
. J D
Shiffer, Vice President Nuclear Pover Generation c/o Nuclear Power Generation, Licensing P acific Gas and Electric Company 77 Beale Street, Room 1451 S sn Francisco~
CA 94106 PACIFIC GAS AND ELECTRIC C01'iPANYd - DIABLO CANYON NUCLEAR POWER PLANT The following documents concerning our review of the subject facility are transmitted for your information.
Notice of Receipt of Application, 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 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 digdidi tl d
d dd d
.d d~d ddd Exemption,. dated Construction Permit No.
CPPR-
, Amendment No.
dated Facility Operating LicenseNo, Amendment No.
dated Order Extending Construction Completion Date, dated Monthly Operating Report for transmitted by letter dated Annual/Semi-Annual Report-transmitted by letter dated Encl osures:
As stated Office of Nuclear Reactor Regulation CC:
See next page OFFICE/
SURNAME)
DATE/
DRSP JPD5 JLee
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/13/87
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NRC FORM 318 110/80) NRCM 0240 OFFICIAL RECORD COPY
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