ML20079L205

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Informs of Results of SEP Review.Rept Will Be Published & Distributed by 821230
ML20079L205
Person / Time
Site: 05000000, Oyster Creek
Issue date: 12/21/1982
From: Dircks W
NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
To:
Shared Package
ML20079F227 List:
References
FOIA-83-363, TASK-PII, TASK-SE SECY-82-496, NUDOCS 8401250370
Download: ML20079L205 (66)
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.p..... %% w ._ s.pm,y. p,,.,,-.g j 4 <? p * *** % II W S Cecember 21. 1982 (b [ SECY-82 496 %,"..W.../ POUCY ISSUE (Information) FOR: The Comissicners FROM: William J. Dircks Executive Director for Operations

SUBJECT:

SYSTEMATIC EVALUATION PROGRAM (SEP), OYSTER CREEK INTEGRATED SAFETY ASSESSMENT PURPOSE: Yo inform the Comissicn en the results of the Systematic Evaluation Program review for the Oyster Creek facility. BACKGROUND The Oyster Creek Integrated Safety Assessment is the AND third review to be ecmoleted as part of the Systematic DISUESION: Evaluation Program. The previous two Integrated Safety Assessment Reports, Palisades and R. E. Ginna were forwarded to the Comission by SE".182-378 dated September 15, 1982 and by SECY 82 403, dated October 1,1982, respectively. The draft Oyster Creek Integrated Plant Safety Assessment Report (IPSAR) was issued as HUREG-0822.. in September 1982. The findings in this repcrt were discussed with the ACRS on November 4,1982. The ACRS reported their evaluation of this assessment to the Comission en November 9,1982. For the most part, that letter supported the staff positien en all issues and comented en points of disagreement between the staff and the licensee. Since that time, the licensee has modified their positions en all point: cf disagree-ment. There is new acceptable resolution en all of these items. Therefers, there are no areas where a disagreement exists between the staff and the licensee. ~ The ACRS letter and the staff's response along with the licensee's cemitment letter are included as Enclosure 1. CONTACT: Robert Fell 49-28923 8401250370 831018 PDR FOIA WEISSS3-363 PDR_

g .y 9 sr The Comissioners, As a result of the Systematic Evaluation Program, plant safety at the Oyster Creek generating station his been improved and a docmented description of the extent to which the plant conforms to current licensing requirements has been provided. While a naber of safety improvements remain to be implemented or further evaluated, the staff has concluded that an adequata basis for continued operation exists. SAFETY IMPROVEMENTS: Safety improvements applied to the Oyster Creek plant fall into three categories: hardware modifications; procedur1s and technical specification changes; and furt ser engineering evaluations to define any necessary modifications. For the casas of confirming analysis, the acceptance criteria have been defined. Some examples of these improvements are listed below: (1) Improve de power availability by installing battery ~ and circuit breaker status alarms. s (2) Improve availability of emeisency power to vital systems by appropriata modifications to prevent automatic transfer of faulted loads. (3) Japroved protection against pipe break'ty install-ing appropriate leakage detection systams (both inside and outside contairaent). (4) Idprove reliability and availability for shutdown cooling under adverse weather conditions (e.g., hurricane inducad flood) by requiring a minimum l quantity of water in the condensate storage tank. (5) Improve assurance that accident related offsita doses are as low as practical by revising the Technical Specifications to incorporata lower reactor coolant iodine limits. (6) Improve availability.of raactor protection l systems by revising the Technical Specifications to include surveillance testing of the emergency condenser actuation components and'1ogic channels. l l ~ N s l l m

. ~... s. .r i The Comissimers ' (7) Imoreve availability of ESF motor operated valves by evaluating thermal overload settings or bypasses for ESF valves and make appropridta sodifications as required. (8)~ PrGvide assurance that the low pressure portion of the reactor water. cleanup systes will not he overpresrurized by demonstrating that the relief capacity of tne safety valve is adequata to protect the low pressure components and does not creata unacceptable loss of coolant conditions. Modify if adequata protection is not provided. 1 i (9) Provide assurance that the containment spray and core spray pump motor can operata during accident conditions with a loss of ventilation; upgrade pump setors or v?ntilation systans if required. (10) Provide assurance that the plant is adequataly protected from tornado wind loads and tornado induced missiles by evaluating the consequences of loads,and missile impact on various building and components, and provide at least one train of protected equipment to assure an ability to safely shutdown the plant. CONSULTANT REPORTS: As indicated in the discussion with the Consission on October 22, 1981, in order to provide an additional level 3f perspective, the Oystar Creek IPSAR was also ruviewad by four independent consultants to the staff i whose coments were available to the ACRS; these consents are presented in Enclosure 2. The consultants and ACRS generally agreed with the approach being taken by the staff. Staff responses to specific comments ande by the consultants are also presented in RESOURCES-The NRC staff review of Oyster Creek involved 7.4 Professional Staff Years of effort and approximately $685,000 in Technical Assistance. SCHEDULING: The staff intends to publish and distribute the Oyster Creek Final Integrated Plant Safety Assessmer' Report by December 30, 1982. ~

v 4 The Commissioners -4

  • POL CONVERSION:

The Final Integrated Plant Safety Assessment Report and Suppleent addressing resolution of follow-up requirments will be one of the bases in considering the conver'.on of the Provisional Operating License i (POL) for Oyster Creek to a Full Tenn Operating - License. The staff is preparing a paper describing the proposed plan for conversion of seven Provisional Operating Licenses. The paper will be submitted to the Commission by the and of 1982. IMPt.EMENTATION: By letters dated September 15, 1982 and Novaber 29, 1982, the licensas provided a schedule for completing i the analyses required for several topics and for completing the modifications resulting from the SEP ~ reviews. We intend to issue an order confirming these commitaants. Upon issuance of the final integrated assesment report, the Director of the Office of Nuclear Reactor Regulation will notify the licensee by letter that ~ the Systematic Evaluation Program for Oyster Craek has beari compieted. Pursuant to 10 CFR 50.71(a)(3)(ii), the licenses will then be required, within 24 months, to " file a complete FSAR which is up to data as a maximum of six months prior to the date of filing the revision." This updated FSAR will incorporate the plant modifications, procedure revision (where procedures l were required'rather than hardware modification) and analyses completed as a result of the SEP review. 4 William. Di re s Executive Director for Operations

Enclosures:

1. ACRS letter on Oyster Creek, SEP dated November 9, 1982 and Staff Response (letter from. H. R. Denton ? i to Dr. Paul $hewson, dated December 2,1982). l 2. Consultants' Ccaments on Draft NUREG-0822 and Staff j! Responses to Consultants' Comments. i 3. Proposed Final NUREG-0822 (less appendices). (Commissioners, OGC, OPE & SECY only). s ~

~~ e .o .e e -- i ENCLOSURE 1 t. 4*""**% UNITED sTATss 7. NUCt.fAR REGULATORY COMMISSION { I j/ AovlscRY COMMITTIE CN H'. ACTOR sAFIGUARos wason= aron, n. c.nossa November 9,1982 Honorable Nunzio J. Palladino Chairman U.S. Nuclear Regulacory Commission - -~ Washington, D.C. 20555 e. Dear Dr.<Palladino; i

SUBJECT:

ACRS REPORT ON THE NRC SYSTEMATIC EVALUATION PROGRAM REVIEW 0F THE 0YSTER CREEX NUCLEAR GENERATING STATION A i During its 271st meeting, November 4-5, 1982, the ACRS reviewed the results of the Systematic Evaluation Program (SEP), Phase II, as it has been applied i to the Oyster Creek Nuclear Geaerating Station. These matters were discussed also during a Subcommittee meeting in Washington, D.C. on October 26, 1982. e During our review, we had the benefit of discussions *with representatives of the General Public Utilities Nuclear Corporation, the Jersey Central Power & Light Company (Licensee),, and the NRC Staff. We also had the benefit of the documents listed below. TMs is our third review of t'he appiteation of Phase II of the SEP. We reported to you on our reviews of the Palisades and R. E. Ginna plants in ~ lettars dated May 11, 1982 and August 18, 1982, respectively. The first report included comments also on the objectives of the SEP and the extent to which they have been achieved.' Our review of the SEP in relation to the Oyster Creek plant has led to no changes in our previous findings regarding the program as reported in our letter on the Pal.isades plant. ~ The remainder of this letter relates specifically to' the SEP review of the Oyster Creek plant. Although the Oyster Creek. plant is the first boiling water reactor (BWR) to be reviewed under the SEP, the findings by the NRC Staff regarding the number and nature of topics for which the plant did not meet current criteria were not markedly different from those for the Palisades and Ginna plants. A large number of these topics related to the adequacy of the design to resist extreme external phenomena (flooding, tornado, earth' quake), and most of the remaining topics related to balance-of-plant items, or items of a generic. i nature not specific to SWRs. i. Of the.137 topics to be addressed by the SEP, 30 were not applicable to the Oyster Creek plant, and 24 were deleted because they were being reviewed i generically under either the Unresolved Safety Issues (USI) program or the i TMI Action Plan. Of the 83 topics addressed in the Oyr'er Creek review, 38 e t a w +ee.

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  • Honorable Nun::io J. Palladino November 9,1982 were found to meet current MC criteria, and 5 were found to be acceptable on l

another defined basis. We have reviewed the assessments and conclusions of the MC Staff relating to these topics and have found them appropriate. t l For all or parts of the remaining 38 SEP topics, the Oyster Creek plant -~ was found not to meet current criteria. These topics were addressed by the lategrated plant Safety Assessment, and various resolutions have been j proposed. i Tne Inttgrated Assessment has not yet been empleted for all or parts of 13 topics, for whic\\ the Licensee has agreed to provide the results of studies, analyses, and evaluations needed by the MC Staff for its assess-ments and decisions. All of these topics are of such a nature that hard-were backfits may be required by the MC Staff for their resolution. The Staff's assessments will be provided in a supplemental report that will. be available for review in connection with the application for a full-term operatino license (FTOL) for the Oyster Creek plant. For all or parts of 10 topics included in the Integrated Assessment, the MC Staff concluded that no backfit is required. We concur. For the reaining topics for which the assessment has been c::mpleted, the MC Staff requires the. addition or modification of structures or equipment in about half of the cases, and the development or modification of procedures or. Technical Specifications in tbp other half. The Licensee does not agre with the MC Staff's requirements for three of the hardware backfits, two of which relate to leakage detection systems, and for five of the required l procedural backfits, all of which relate to the Technical Specifications. Our comments on these areas of disagreement are given below. In connection with Topic III-4.A. Tornado Missiles, the MC Staff's concern is that all of the canponents that could be used for shutdown heat removal could be disabled, by multiple missiles transported by a single tornado. The MC Staff requirement is that at least one system capable of shutdown heat removal should be protected against tornado missiles. The Licensee believes that the total loss of shutdown heat removal capability as a result of multiple missile strikes is of such low probability.that no protection is needed. We agree that this is a very low probability event, but we do not believe that the probability has been quantified with any significant degree i of certainty. Further, we recognize the importance of having at least, pne shutdown heat removal system asailable following a tornado, or other extreme environmental event. We recamend therefore that one such system be pro-j tacted against tornado missiles (and other possible effects of high winds, such as sandstanus) unless the cost of such protection clearly outweighs a the reduction in risk. ~ 4 l I For Topic III-5.B. Pipe Break Outside Containment, the MC Staff requires an automatic local leakage detection system for t!.e isolation condenser piping, I, ~ i i

.s . q. 4 l Honorable Nunzio J. Palladino November 9,1982 l which is lagged and is outside of containment. The system should be capable of detecting leaks from stable cracks before they grow to be too large. The detectable leak rate is based on an analysis of tight cracks whose length is two to four t%es the wall thickness. The Licensee contends that the leak rate corresponding to such a crack will be large enough that it can be ~ If they cari show this to the NRC Staff's " detected by visual inspection. satisfaction, we feel such an approach is simple and reliable. If they cannot, an automatic leak detection system would be a more delicate but' acceptable approach. .s Topic V-5, Reactor Coolant Pressure Boundary Leakage Detaction, relates to the requirement for a reliable system to detect leakagg, insida the contain-mert with a sensitivity adequate to provide aarly warning so that timely actions can be taken to preclude a pipe break. The Licensee believes that the existing system, utilizing the containment stanp, is satisfactory. We believe that this matter should be resolved in a manner satisfactory to the NRC Staff. In connection with Topics V-5, VI-7.A.3 and. VI-10.A, the NRC Staff requires that certain limiting conditions of operation, and surveillance or test requirements, be added to the Tee:hnical Specifications for the Oyster Creek plant. We concur. Topics XV-16 and XV-18 relate to the calculated radiological consequences for rartain design basis sccidents; thyroid doses calculated in accordance with' current criteria are considerably in excess of the siting critaria. To correct this situation, the NRC Staff requires that the iodine concentra-tion in the reactor coolant be limited by appropriate changes to the Techni-cal Specifications. We believe that this proposal is acceptable. As was the case for the Palisades and Ginna plants, a plant-specific proba-bilistic risk assessment (PRA) was not available for the Oyster Creek plant. Because a plant-specific PRA was not available, the NRC Staff utilized in its Integrated Assessment the results of the Millstone Unit 1 PRA developed as part of the Interim Reliability Evaluation Program (IREP), suitably mcdified and 11terpreted to reflect the differences between the two plants. The PRA study for Oyster Creek addressed 20 of the topics included in the Integrated Assessment, a somewhat greater number than for either Palisades or Ginna. l However, because the Millstone IREP did not include extreme external events. topics relating to design criteria for such events could not benefit from,the use of PRA in the Integrated Assessment. Our conclusions regarding the Oyster Creek SEP review $re similar to those i for the Palisades and Ginna plants: 1. The SEP has been carried out in such a manner that the stated objectives have been achieved for the most part for the Oyster Creek plant and { should be achieved for the remaining plantr in Phase II of the Pr.ogram, s 1 ~ w

i I Honorable Nunzio J. Palladino" 4-November 9,1982 2. The actions taken thus far by the NRC Staff in its SEP assessment of I the Oyster Creek plant are acceptable. 3. The ACRS will defer its review of the FTOL for the Oyster Creek Nuclear Generating Station until the NRC Staff has completed its actions on the remaining SEP topics and the USI and TMI Action Plan items. Sincerely, P. Shewson Chairman Referencas: l

u. S. Nuclear Regulatory Commission Draft Report, " Integrated Plant Safety Assessment, Systema::ic Evaluation Program, Oyster Creek Nuclear Generating Station," NUREG-0822, September 1982.

2. MLC Staff consultancs' reviews of the Oyster Creek Integrated Plant Safety Assessment Report consisting of consultant reports from H. S. Ishin, I. Zudans, J. M. Hendrie - and S. H. Bush, dated October 22, October 25, October 21 and October 20, 1982, respectively. 3. U.S. Nuclear Regulatory Commission Safety Evaluation Reports, Oyster Creek Systematic Evaluation Program ** Topics, Volumes 1 through 3, dited October 1982. a O e ~ e l I .l 9 i l e l -m 4 --.m.,m-,,,.-,- --.--,e,,,-. w --,,,,

/o.o**g), UNITID STAT!s 7. NUCt. EAR REGULATORY COMMISSION ~ { 2) wasMmoros. o. c. 2osss p December 2,1982 1 8 Dr.' Paul S. Showson, Ch' airman i Advisory Committee on Reactor Safeguards U. S. Nuclear Regulatory Conurission ~ Washington, D.C. 20555

Dear Dr. Shewson:

In your letter to Chair. nan Palladino dated November 9,1982, the Ata presented its views on the Systematic Evaluation Program Integrated i Assessment Report for the Oystar Creek Nuclear Generating Station. In summen, this letter supported all of the staff's positions except departed from the staff on leakage detection. The purpose for this, mescrandum is to respond to the Consittee's coments and describe the i subsequent resolution on leakage detection and other issues of staff-l licensee disagreement. The licensee has agreed 'to provide at least one equipment train which is protected against tornado missiles. The Itcensee has verbally . advised the staff that they are evaluating the installation of a portable pump and hose connections to provide cooling water to the emergency condenser. A schedule for this modification is forthcoming. The staff concludes that such a proposal would be acceptable, subject to corifirma-tion that the equipment and water supplie.s are in protected areas. For the emeroency condenser steam ifne piping outside containment, the ACRS suggestad that visual -inspection for leakage should be acceptable i if the licensee can demonstrate to the staff that a crack sufficient in size to produce visable leakage is stable. The licensee (GPU) is proceed-ing along this path and the staff will act on their proposal scheduled to be submitted in Februan 1983. If the limiting leakage rata cannot support visual inspection, then some form of automatic leakage detection l would be required ecsuunsurate with the limiting leakage rata. l For reactor coolant pressure beundan leakage detection inside containment, the ACRS judged that this issue should be resolved in a manner acceptable i to the NRC staff. The staff and licensee have reached an agreement on i this matter. The licensee will evaluate the reliability of their existing sump level monitors and atmospheric activity monitors, and modify them, if required, to detect leakage rates from the limiting stable crack. The . l licensee will then propose specific action requirements for the Technical Specifications to respond to changes in leakage or a loss of monitoring capability due to system failure or a seismic event. The staff finds this action acceptable.. The licaensee has also agreed to all the technical specification change issues that were points of disagreement at the ACRS meeting, including i t

/...,,. .... = _ _ _ _.__m _ 2-I the one discussed in the preceding paragraph. The licensee will adopt the General Electric Standard Technical Specification limits for reactor coolant iodine. However, the sampling frequency and corresponding actions 'e will be developed on a plant-specific basis as a function of the samoling -technique and plant operational characteristics. The licensee has agreed to incorporate reactor protection system surveillance testing requirements into their Technical Specifications for the emergency condenser actuation couponents and logic channels and testing of the reactor made switch, high drywell pressure instrumentation and manual start and timing relays, all of which interface or are part of the r'eactor trip system. The staff will revise draft NUREG-0822 to reflect the agreements reached between the staff and the licensee and respond to the recoamendations from the Consrittee and the staff's consultante. I ? Sincerely. ^ Original Egned By Harold R. Der. ton, Director Office of Nuclear Reactor Regulation e e e e 9 a 5 e 1 m-g+- 3-v----v -yi-w-r--g, ,,-,t-i- ,, - -.ww

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~~ ~ ~ ~ ].; ~ ~ ". [Alclast.4t::, ! f3'#1J Muelear f 3-j 'QQ $[ 100 Intercace Parkway Psrsccany. New Jersey 07054 201 263-6500 TELEX 136 482 Writer's Direct Dial Nummer: November 29, 1982 i i + Mr. Dennis M. Crutchfield, Chief Operating Reactors Branch f5 Division of 1.icensing U. S. Nuclear Regulatory Commission Washington, DC 20555

Dear Mr. Crutchfield:

.+ SUBJICT: Oyster Creek Nuclear Gdaerating Station Docket No. 50-219

  • SEP Integrated Assessment Cur recent letter dated November 16, 1982, provided our schedules for equipment modifications and additions resulting from SEP integrated assessment. The letter also provided GPU Nuclear Corporation (GPUNC) positions concerning resolutions of some of the SEP topics proposed by

( the NRC staff with which we disagrcel. Subsequently, the NRC staff requested a meeting for further discussion of the schedules and topic resolutions in order to reach agreement between GPUNC and the NRC staff. That meeting was held on

  • November 18, 1982 and we were able to reach ageoment on all items.

ATTACEMENT I to this letter lists schedule and resolution agreed at our meeting for each of the topics. e f n / i l t t ' ~ GPU Nuclear is a part of trte General Public utilities System

i._. -- . _...._...m._ ._.._....__......_.i } .; Mr. D. M. Crutchfield Pags 2 November 29, 198: We are also transmitting the revised integrated assessment s summary table (Table 4.1 of dr'af t NURIC 0822) as ATTACHMENT II. The revised table reflects the agreements reached during our November 18, 1982, meeting with the NRC staff. The revised portions of the table are i . indicated by double asterisks (**). Very truly yours, i ~ LaclaC

7. 1. Clark Executive Vice President I

bif Attachments cc: Ronald C. Haynes,. Administrator Region i U. S. Nucisar Regulatory Comunission 631 Farh Avenue Ring of Prussia, NJ 19406 NRC Resident Inspector Oyster Creek Nuclear Cenarating Station Yorked River, NJ 08731 ^ t t t rr l i l l 1 4 i l

....i.'.. T.%.... _ u ~.~.~ e 7 ATTACEMENT I SIP TOPIC SCHICULES AND 2ESOLUTIONS AC2EID ON NOVIMBER 18,1982 BETUgEE CPUNC AND THE NRC STaiy TOPIC NO. III-33 Pipe Break Outside Containment

  • As stated in our November 16', 1982 letter, CPUNC will prepara and transmit.a report to the'NRC by yebruary 25, 1983, which predicts crack

^ growth rate l'a the Oyser Crensk isolation condenser lines. Resed on our findings G7UNC will propose an appropriate leakage detection method with adequate sensitivity. Our preliminary evaluation indicates the extremely slow growth race will allow the use of a crack with suf ficiently large leak rate for visual and audio detection TOPIC NO. V-5 Reactor Coolant Pressure Boundary Leskate Detection GPURC will perform a " leak before break" analysis for the most Limiting piping to justify the.wnsitivity of the current leak detection systems available. in the containment sump, Results of the analysis will be transmitted to the NRC by June 30, 1983. CPUNC will also propose Technical Specification change before Cycle 10.startup to include operability requirement of the, leakage detection systems inside containment. CPUNC'will prepare action statements in Technical Specifications. TOPIC NO. VIII-33 DC Power System Eus Voltage Monitoring and I Annunciation Our November 16, 1982 letter stated that the battery status alarms.will be installed during Cycle 11 refueling outsge due to the number of modifications alrudy planned in the control room during the forthcoming-(Cycle 10) outage. However, the NRC staff requested an intermediate resolution of this issue since FRA performed by the NEC consultant rated. this issue of high risk importance. GPUNC will provide an intf;rmediate resolution which consists of periodic inspections of the safety related bettery systems such as verification of battery breakar l position, battery current, charger current, etc. Plant inspection l procedur.e will be revised to include the pariodic inspection of the battery systems before cycle 10 starup. The intermediate resolution will ^ be in effect until the installation of the battery status alarms during cycle 11 outage. (- i TOPIC No. V-12A Water Purity of EUR Primary Coolant ' Our November 16.1982 letter to the NRC stated that CPUNC vill revise Oyster Creek Technical Specifications after examinics our C-c*.e 10 i experience. However, during our meeting on November 18, 1982, the NRC staff requested C7UNC to revise Technical Specifications before Cycle 10 startup. G7UNC agreed to revise Technical Specifications prior to Cycle 10 startup incorporating Reg. Guide 1.36 ' limits for conductivity and S .a e e.. .....en- =. 4-. ~ r eiw +w' vr -*.rw+ =-w-re y-- .--,e ,.w,-wyi-g.m in eime-w-,. - - - - -,, w w- +. -, mw--c-- wwy--wwm--.q-,wvew,,,y+.,-g.

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- ~ _.. -.. - -.. ~....- n. f 4 Action st'tements will be proposed by CPUNC. chlori 8e concentration. a The NRC staff concluded in'the meeting that an issuance of a special report rather than an LEE by CPUNC would be acceptable following a violation of f.he Ees. Guide limit for conductivity. The existing linie for conductivity will still remain as an upper bound for the plant operation. TOPIC NO. XV-16, IV-18 Radiological Consequences of Failure of Small Lines Carrying Primary coolant and Main Steam Lines Outside Containment 1 l GPUNC will revise Oyster Cnek Technical Specification adopting General Electric Standard Technical Specifications for iodine linits j before Cycle 10 scarcup. G?UNC will propose action statements. . TOPIC NO. III-4A Tornado Missiles i GPUNC will provide a portable pump and hoses fer shutdown heat j reactal during a posentated tornado missile incident. The pump and hoses a will be housed in the. area of Eesctor Building which is protected from ths tornado missilas and flooding (the pump will also be used as a i redundant pump to the existing condensate transfer pump during a postulated mazimum hurricane flood -See SEP Topic II-33.in ATTACHMZNT + II).- S'ince placing of the pump and hoses is not outage related, CPUNC will complace the project by the end of 1984. The project also includa revising the plant emergency procedure for tornado incidents. "CPIC.40. "I-7. A. 3 Emertene-r Core Cooling Svsta= Ac:ua:i:n 3vsta= . TOPIC NO. VI-10A Testing of Reactor Trio Systes and ESF. Including Responsa Time Testing Before Cycle 10 startup CPUNC will revis e Technical Specifications to include the emergency condenser logic testing and the reactor trip system testing in accordance with the instruction provided in our current surveillance procedures for both systems. l 1 t l l t i I 1 1 I i

'n-y labl2 4.1 litsgrt.te:el assensseiet suma ry I. sa g* 3 n lech. Spec. 'I y 5fP g aspellfications Cie ple-a lepic Section respelred free untilt Lisensee t iesa PRA No. Ilo. Iltle 5fP review respelrennent s . agrees date rating' sn I 5 11-3.8, 4.l(I) Condentata Iransfer Pamp iso. See SEP Topic No. III-4A** pes e 12/a4** l 11-3.3.1, Pouer Section No. 4.6.-l 1 ~ l ll-3.C 4.l(2) floodinglevel Procedures lia None i 4.l(3) Canal Water level lia ~ lustall aulusiatic water Ve> Cycle kl I lastrimentation t level lustroes:nt.ation in j the intake canal. I 4.l(4) Isolation Condenser - llo Besonstrate minim a quantity Yes Cycle X flooding of waler anslutaisied in l i renalentate storage tank ~j sufficient for long-tera e $i coolind and inclemle minleun inventury la pl.nat proceennes. { 4.l(5) Low Water level 58mendown No skuie ~ ~ 4.l(6) tharricane flooding of-He Revise ca repnecy procedures Yes Cycle X Pumps to filenilly alternate water souries Awl Ilenaliaths should low clevallon tasaps be f leo.leil. j 4.l(l) flooding Elevation . tle Evalulta a.onse.p.ences of Yes 2/83** offu.as i.eslidle.1 flooding and 1 cenfiras all other entrance i 4 levels.dasve 2:1.5 st. I 1 R 4 + ' See foetnotes at end of table. i J .I j

y i m late 4: 1 (Cs.ntinue.1) g l .i ,t s t j n Z SEP lech. Spec. i 'iopic Sectlen modiIIcatlens so Compta-No. Ilo. Iltle respaired from Sackfit Licensee Lion PAA sa $Er review resguireiunts S agrees date rating' Il-3.C 4.1(8) Groundwater Elevation lia See"ltce.4.4(2). ?{ i . 4.l(9) Roof Drales leu install nuppers la the Yes Cycle X reacter 84 sliding and tierblas 4 me. l ;. - .l bulIding par.apels. i til-1 4.2 i Classification of Struc-No l1 tures. fooposunts, and (valuate deslun of 2Pecified Yes le CFR Systems comanneents on.s 5.eap11ag 50.71 bawls, sporade if necessary. (e)(3)(ll) i 4 anel A=.unent classification la F5 Alt agal.ste, \\ s lil-2 4.3.1, teacter Sullding Steel No Structure Above the Analyse and identify any Yes 12/82** l i Operating fic.ar needed og grading al reactor j buildian upper steel I structure for wismi leads. 4.3.2 Ventilation Stack Nac Analyse and identify any Yes 12/42** i needed us. grading of wantila-tien St.:.L for wind laads.

4. 3. 3 filects of failure of elo Nonselselc Category I Analyse turbine lenilding Yes f./81 4

Structures capacity for wisul loads. evalu. ele conteepn nces of i failurer de=I 14 nlify any needeel a.dr.dism,s. n i 4.3.4 Components Not faclosed No lione i in Quallfled $tructures 1 t i 3 4.3.5.. Interior Hasonry Walls No Is sa 4 k 't See footnotes at end of table. I i. i t 4

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y -g ~ Tetti. Spec. ,n SEP modifications (Icensee tien PSA j Cangele- ..g repic Sectis.4 reresired fran Sectilt n lie. fle. Iltle SIP review . reeluireements agrees date rating i 4 4 an lil-2 4.3.6 Boef Becks lin Prowlele analysts el reacter Yes 12/82** hullding rse.f. ~ ~ Analyse c.apacity of turbine Yes 6/83 bullJlsed conf In withstand wised Ia.s.ls. ~ 1

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4. 3. 8 lead Cae6fnations lie Sedeelt analyses deen.n-Y.* 8
  • 6/83 "

i s strating wised le.nl were l 8 properly cas&ined with other i; j / speciiled leads. 4.3.9 5 ell and foundation He thune Capacities I- ~ i Bil-3.A 4.4(1) N9ilrestatic toads -Ne kne (Comblaation) 4 4.4(2) liyalrostatic loads lie Evaluate i, Inert shnallen Yes 4/83 j (Shoit-Duration) laydrost.itic leads on and floatalian i.elestial of structines essential to sale sleul. bene la ca.nlinictlan-with iI. lis.g emergency prece.bu e. Iliene 4. 3(6))., 4.4(3) Releia-Grade Penetralien b ika.e ) ffooding ) \\ i i t l i g i i 1 i l

  • * - * * ' ~

m s e. m s ' - - "h, j Tech. Spec. .Stp 'I n matificatleans i I g Topic Secties Caneele-e lie. No. Iltle respired from Battlit Licensee Llose P24 t W 5(r review respirem.:nt. . agrees stat e rating M lit-3.C 4.5.1 Intake and Discharge Canals lia IU Ifwm, 4.5.2 Intake Structure Trash No l Sacks and latake Screens Fernalle existing laspec-Yes* Cycle X** tien prm.tice as p. ort of i stilIL larwever nr laservice inspes.ll.au list) pr.cedures instil w. iter level medist-l catlen la complete i (lien 4.l(3)). 8' 4.5.3 Seef' Drains Na See item 4.l(2) 4.5.4 Inspection Program No I pavelop. pal implem. rut a Yes Cycle X formal inspectima program ,I e for w. ster cwitral structures. III-4.A 4.6.1 Emergency Olesel lie Generators and fuel Analyse potenti.el for and Yes 3/83 4 Gil Day Iank constepwnces es t.sraado-i alssile d.. mage at the diesel

  1. j generator Semitellag.

~i 4.6.2 Nechan'ical Equipment lia ? Access Area Evaluate the potential for Yes I/81 aent conteepwnces al tornade-i } missile impact la the reactor j j luellells.g.sccess sluer region and I.lentily any necessary {' correceive attle.ns. 4.6.3 Control Room, teactor Ne llane Sullding, and Turbine sulldinglleating, j i i. . ventilating, and Air j Conditioning (llVAC) Systems i i s, i See footnotes at eeul'of talsle. i e \\l l

e 7 Y e ista 1.3 (l'ontlaum.8). I! y

l rs lech. Spec.

86 lepic Sectlen tangele-I 2 SEP mellfications reepstred from pactIil Licensee t is e PSA lie. Ole. Iltle m 1(P review repelrtw.ets agrees date rating 5 PreviNoretectionter Yes* I 3 2/84** Ill-4.A 4.6.4 Condensate Storage Iank, Na 1erus Water Storage Iank, sufficient syste=> and l t f t, and Service Water and cangu sm mes to em>=re a safe l Emergency Service-sl=eteham in time ervest of Water Pimps dama se isum tern.s.l.a missiles.4 flooding Ill-4.8 4.7 lurbine Missiles lie luspes.t lue1.isa.eA.i propose Yes 12/b4 leapectisen frespa:sicy based on resells. .hettily amesillarlos program Yes 2/83* for m.ala ste.am.n=4 rebeat control v.alves.. lil-4.8 4.8.1 Truck implesien He Nimie l 4.8.2 Aircraftliarards No fuslu.ste potential for er Yes 3/s) l l casese pacae.es el alscraft - lag ae t. lil-5.A 4.9(1) Cascading Pipe Breats No' see lleu 4.16. 4.9(2) Jet lapingement Iifacts lie Ene 8 4.9(3) Drywell Penetration He bne see footnotes at end of table. J \\. 9 .I i t I

.Y p f.al le 9.1 (r. t i.ne..ll .g 'l m i rs _.q 2 SEP 'lesh. !pec: s.nlit icat ionis l lepic Sectlen ressuireil irsie II.n i s e s Licensee tien PAA Caeq.le-a No. No. Iltle Str review n eiriir.s :nl3 agr.res slate ratir.g. '{ '8 lll-5.8 4.10 0 ) LOCA outside Containment iso - Evalu..se an.1 flentify any Yes 3/u3 l } sierdeel asperading of the -t mala steam and reacter l wate'r cleanup (RWCU) piplag outside containneeni. to .l l preclu.h aos emaiselatable les est evalslele regitalament, i , i a 4.18(2) Emergency Condenser Na I.v.alu. ate ased Islent ify modi-Ye:. Cycle X H l Isolation i fic.etion tu previele leakage eletc.:lium to ensure flaws u nslia les detta.tud f efore I: 3 .+ pipe Isreak eccurs. l' al. Ill-6 4.1)(1) piping Systems Ne' Analyse on a s.imgsling ,Yes Cycle X 8-* leasts aimi verily adequacy ut seggwert slesl:Isig for the sel>=lt resist.u ce of specilled pipihu systems. l 4.11(2) Hechaalcal Equipment lie lii n.si:.trate th.st ti.e control fe. 12/82* real alrive spt.-es.inI vesss i Intern.als have sullicleat i e:.sp.= ily la resial a safe 't i .l.nl.l..use c.srth.pe.ste or take

s. ors. s I she att laws.

1 i 4.1)(3) ' Electrical Equipment No Neev.slu.ite 486es-V switch-Yes Cycle 2 go. ar p.uuri sm h.or.ag.: asul sIart elitan.:.t r. ale, sus.a e,ampilag su ami. l o ls, adespeay el electrl-j c.sl p es.c1.'s.sqq=.e I s. ~.~. See featnotes at end of table. i I I

j o- . - - t labla ' (Caetlem el ,{s sa t&- q -l lech. Spec. Q $(p sia.lille.es imes I:; Ispic Sectles Ca.agele-re.pilre.1 f re.a. B.= I I is a1. *,cw Lina FRA w No. Ms. Iltle ' Sir revicu r..reis.y nts ,agre :s it.et e ratlag m Ill-6 4.11(4) Ability of Safety- - tse n. i Related Electrical Equipment to function. l 4.11(5) Qualification of Cable Ne frays Fr widir plan to luplement Yes Cycle M 4 !l re> wits of 5tr Owesers i i lirege Fr.agr.ma ma a plant-spes.itic basis. Ill-7.8 4.12 Desiga Codes, Design Ne j Criteria, lead Iv;slaats ade.pucy of Vr> 6/a3 Combinallons and Reacter e.rluiad desi ta criteria t i u 4 s.ingella.Ile.swis for i. Cavity Design Criteria t - H.illeil 38 rem.tural i us e I e.L 3. to Ill-8.A 4.13 toose-Parts Nealterlag Na i i Is.ine i and fare Sarrel tow Vibration Healtering i Ill-lo.A 4.14(1) 1hermal Overload Bypass 18e p i I tv.ilu.ete thermal averload Yes l./s nwa for es.gineeret 12/82** II'dI'"I 1 t..st.sy. feats <> (456) valves. j 4.14(2) Haga5 tic 1 rip Breakers W hs.- 1 I ly-2 4.15 Reactivity Control Systans, h II.e : 3" laciudlag fianctlanal tow Design and Protection i A alast Single failures 0 i See footnotes at end of table. ) i ~ y i. l

u ~ Iale*'4.1(One.Linmed) l se . 1 lecle. Spec. { ~ SEP i moellficalloses Casmaile-i lepic Secties regelred from leadlit Licensee tfan PAA' w No. lie. Title 5tP review respelrrments agrees d.et e rating e,. l Pd V-5 4.16.1 Airborne Particulate and H. Hake those monitors yes. Cycle X** t w

l j,

e Gaseous Radioactivity operational Honitoring i i i lople lil-5.A analysis, i! l 4.16.2 Operability Sequirements Yes las.lete trat.epe detectlen Yes** 17/82 l syse.:n lleitinia e.ouills. leas Inr everatlose in iethnical Spes ilicatlens. t 4.16.3 latersystem teaka:le Ile is.un, , us Inventory Salances laun.. t 4.16.4 Reactor Coolant Nu e i 6s l : Vr5 4.17 Reacter Vessel Integrity Ne bd. sell a plan ser the Yen 12/s2** + ! 'i I material surveillance I rapsules. i V-10.R 4.18 Residual lleat Removal .IIe Newli.w amt upisraele, if Yes Re f' ore

  • tow

} ! systee Reliability mete:.ssry, shuldinne Cycle 10

I presclures 16. Specify Startup j

't altera.ste sources of j wales for primi.ery aeul seceol.asy mateny, with j $, particul.or atlentlen to enternas aves.Is. i I V-il.A 4.19 Require,ents for Isolation No l ,, of liigh ' and low-Pressure Seminisir. ele actief capacity Yes I/83 low" j Sy5 Lens and er.a.a.d.le e unwquencIrs, l t i or fel.:s.: ea r. rrei t ive acLlaa Ie pr.'gl.. 3.9 38 sy, tem. N See l..tnotes.t end., t le. i ~ i ?

3,..., 4, g gg:,,,e 3,,,,.4g ,,. q in ( 'l-L{ leali. Spec. [ 5fP meillIicallann. lepic Section Comple-O reepstred frasi SacLif t l iceer.ee tives PRA F Ho. Ilo. Iltle S(P review re.psiren.ent s brees date rating e,. IU V-12.A 4.24 llater Pierity of tidt Yes ligelement pragwesed procedure Yese Cycle X" Prisary Coolant on. se sity les:hnical Specifi-callesen to 1.e consistent. VI-L 4.21.1 Organic Haterials No Inspert asal repair, if Ves Cycle 2 e [ nes e.t..ary, drywell coallags 86 me. amis retual tsee terns. 4.21.2 Postacci' dent Chemistry No. is.ne.: VI-4 4.22.1 tected-Closed Valves No Provide atlee.l.sle to.laplement Ve> 11/82 tow pleysic.at locLipse devices to p en aere valves are not p In.s.lverteat ly aspessed. ~ 4.22.2 Seeste h aual Valves lio Provide leataspe detectica Yes. Gcle X", i an.l.18 acce:.sary, relocate I. the oper.stisuj station for ,t lie lallwa valves in the cm.t.alsenes.1 spray and core spr y systems. 4.22.3 Valve,locallon h Nwee 4.22.4 ' Instrument times W Hyn.. 4.22.5 Valve locallon and lypq h kn.: ~ 4.22.6 Semote llanual Valves W lim, VI-7.A.3 4.2)..

  • Imergehy Core Coollag Yes incisoli es.:sapreecy condedser Yes" Gycle X" -

System Actuation System lo.,ilt t. s in. in the letlup. I tow isit.atleas. t Q See footnotes at'end of tal.le. e i .l i (

l a.mie 1. a t Lant ionai.8) --[ t U g -s hch. Spec. I r3 Sff' anidisications g 2 . Topic Sectica Cnegele- .'i reipstred from Nactfil Lici m-se-tien PAA I l so Na. No. Iltle w Sil' review respairem nts ' agree-elate rallas l y VI-7.A.4 4.24 Core Spray Herale No saune l a Effectiveness VI-7.C.1 4.25(1) - AC Automalir Sus Transfers No tvalu.eser the esimalug Yes* 12/82 lledium i 3 I antin.alle ines transfers i ami liktalify currective actisms la veesuev taulted leads 6#smid seul lee 4 I ranslea r.r.l. 4.25(2) DC Automatic Sus Transfers Nu laune VI-io. A 4.26.1 Response-ilme Testing No tiene t i Law i 4.26.2 lastruneatation for Reacter No 4 {E verily.all safety logic Yes 12/82 Irlp System (RIS) Iastlag cleannels tieel le the reacter amile switcle are tested by .i proce.bere. Yes lacimka logic chaemel Yes** Cycle X** l testimes in fecimical Specifications. 4.26.3 Oval-Channel Testing No asone Vil-1.A 4.27(1) fluiHoelloringisolation ski Perf arm sallurer m. le and Yes 12/m2 tow effee.t.sualysir, la eletermine uleellurr isolatiem elevices are avereire.1.unt I.I.ntify any := c.li et sq.ter.a.lin.s. 4.27(2)* Reactor'ProtectionSystee is.e - (APS) Protective Irly li.:,l li l~ a e,s. Il 4.r..t'ect ion Ves 8:ycle XI i I II r 8 e

3. ens. s sag ply

.u I IM. a Icrl.a e 9 j j j 5ee footnotes at end of table. Ii.i i I. e i l

y leh* 1.1 (Contin nd). .-..i =e -{, malific,ations ':. g ; iact. 5 e.. g SEP Ispic Sectlen Ce= tele-respaired from Baclait licen.ee t i.= PBA tea. Ile. litle 5(P review respelreen nts ve ane w s dat e ratlag D Vil-1.8 4.24 Trip lancertalaty and ' iks i ~ Setpelet Analysis Review lastall analog trip systaa.. Yes Cycle XI Low of Operatlag Data Base Vil-2 4.29 Engineered Safety featesres System Control logic and See item 4.14(l). Law l' Desiga i 1 Vij-3 4.30 Systecs Regeaired for Safe les Sleutilown Frowl.le slaisem leventory Law for casolensate storage tank- . as a water noen ce ser flood-I ing events (Isew 4.l(4)) t aent.ish ality enan-lif equip-ment in caelduun procedures ,g s (item 4.8s). j i vill-2 4.31 Onsite faergency Power lie Hodify.enmaci.itars to Yes Cycle XI tow l 1 Systems (SieselGenerator) confeim la ifl4 5td. l 279-15/6 amt hyp.au two trips (vultage-ang ere reactive asael reverse power) 1 eherines acclelent cessalilleas. Vill-3.8 4.32 OC' Power Systen Bus lie ~ 12 [82 " i Voltage Healterlag ami. tclee.hele. Install.at leesi of Yes liigh Amuneclatloa Specisicd hattery status i i 41ai n,. i Vill-4 4.33 flectrical Penetrations of fie leone Reactor Contalmnent See footnotes at 'emi of Iahle. 1 \\ i i g. I a I

~ ~ ' ~ ~ g,,g, g 4,,, g g,,,,, ;, ",,,,,,, - - - - - - - ~ i a< 17 lecle. Spec.. f, 5fp sudifications j es . impic Sectica Comiele-I- respelred f ee.m ektsit Licem.ee t ie.se ptA

tse, sie.

Iltle itp review ee. pale. uts aggreres d.te rallag i !U IX-5 , 4.34(I) Bastoraties of Ventllatten tio .....s.. tvalu.es.,.m.d revice, if . Yes - 3/is) suce.s ry, the 8.c.s-of-elisile Inn.vr powedures le esesuse th.st restoration ut vraill.etlose systems will seet e.verlma.I tiec ellesels. l 4.34(2) Reactor Sulldlag b bue Ventilation I ~ l 4.34(3) Core Spray and Centalament b Spray Pun,Ventilallen itemenstr.ste sideject pumps Yes 3/81 1 case aper. ele with a less ] t mi v neilaties, or identify curree tIve.se tinsi, as u. 1 seecen.ory. 4.34(4) Battery. Holes-Generator. Ile. Ivalm.ile =Isasts of less of Ves 12/81 1 and Switchgear Ventilatten wentlI times su the subject ro$ n one.1 ides.t ity.say seced..8 sq=pra. ling. r XV-s 4.35 Decrease la feesbeater ' lie ~ ss..ee j' leaperature, lacrsase in feedwater flow, and in-crease la Steam flow anel I Inadvertent Opeatng of a Steam Generator tellet er Safety valve l' XV-16 4.36 Radiological Ceasespeences Yes Iapleatat.3Wil SlasaLard Yes** Cycle X** f er failure of small Lines Carrylag Primary Coolant lechnic.el Spee.leicatloa outside Containment limit s : priatuy toelant e iris,s. 'r E .I i { ^

~ J -e is l.1 (si.e.tlin w.ss g

  • ~

j,,, -s .n SEP le h. Spec. + J lepic .Section meilliIcatisas 1: the. Ilo. Iltle respelred from takela C.=,le- 'o tlonw, Sn PaA iff. review reepsire e.ts eter. es .s s e rating ej xv-is' 4.37 Radiological Conseapsences ef.a Hain Steam time see lice. 4.3Ei. Failure Outside Containment XV-13 4.38 toss-of-Coolant Accidents les Nesulting free Spectrum 8evelop and inalement a Yes cycle II of Postulated Pipe treaks preventive maintenastie t Within the teacter Coolant progree for the mala steam Pressure Soundary isolatiesevalves,orjust6fy existing malatessence based ma operating esperleisce. Sbitresultsofevalua-si Before" t tien Isa luding testing Cycle 10 I, cycg g g emperience. Startup i

  • lligh for other reassas as emplained in the refereined section.

i' 1 NOIES: 4 1 later - the licensee has met yet responded. Cycle refers to the end of a s,secific refiseling enta08 ley care cycle. I ) g J l 5 / N*. .I a

ENCLOSURE 2 f':,,,,g.cg%, UNITIO sTAiss NUCLEAR REGULATORY COMMISSION _y I a 2 wAsniscTou. o.c.zosss November 12. 1982 Occket No. 50-219 j Dr. Spencer H. Bush Dr. Joseph M. Hendrie

  • Dr. Herbert S. Ishin.

~ Dr. Zenons Zudans i Gentlenen: \\

SUBJECT:

' RESPONSE TO SPECIFIC COMMENTS MADE BY NRC STAFF CONSULTANTS l

i REGARDING THE DRAFT INTEGRATED PLANT SAFETY ASSESSMENT REPORT f Ref,ardnces: (1) Letter from Spencer H. Bush to Christopher I. 'Gefmes, dated October 20, 1982. (2) Letter from Joseph M. Hendrie to Christopher I. Grimes, dated October 21, 1982. d (3) Letter from Herbert S. Isbin to Christopher I. Grimes, dated.0ctober 22, 1982. (4) Litter from Zenons 7.udans to Christopher I. Grimas, dated October 25, 1982. Enclosek for your infonnation is the staff's evaluation of specific ~ questions or comments made resulting from your Niview of the. Draft Oyster Creek Integrated Assessment Report. In general, your review su'pported' l the staff's position on the various topics with only a few exceptions'.* iseme. - These exceptions are identified and addressed in the enclosure.. iuations 8 general: comments pertain to the mschanisms for follow up of "evi with p'otential backfit" items; and the mechanisms for integration 'of TMI and USI' items. Regarding this issue, a program methodology is being de-- veloped to follow up the resolution of these open items prior to issuance of a Full-Tenn Operating License. The final report will be revised to. l correct the editorial errors and incorporate some changes in staff posi-tions as a result of comments received both from your review and from the ACRS. i i Sincerely, dd..jLOT- - -01 ~ William T. Russell, Chief Systematic Evaluation Program Branch Division of Licensing

Enclosure:

As stated .J. . ~. _ _ _ = - - = =

r--.

--=--n.-

,g, 1 Dr.' Spencer A. Bush I Battelle Pacific Northwest Laboratory l Fost Office Box 999 Richla~nd, Washington 99352 ~~- = ...e i Dr. Joseph M. Hendrie Department of Nuclear Energy Building '197C Brookhaven National Laboratory Upton, New York 11973 a s a Or. Herbert 5. Ishin 2851 Montery Parkway St. Lcuis Park, Minnesota 55416 oo Dr. Zanons Zudans Franklin Research Center ' Benjar'.. Franklin Parkway Philadelphia, Pennsylvania 19103 i a i T l I I i 9

  • 8 m

? 1 i' . -... ~

i ^ l v. i (1) Comentr (Ishin) s In my judgment, the overall format. used to present the issues and decisions is good, as well as the process of decision making. The concise presentations retain the desired clartty, but may not reflect the balance in the decision making as perceived by the licensee. For example, future reports, and in particular the Supplement, might include interactions of j the licensee with industry-sponsored programs.as well with other ongoing i NRC activities, the initiative displayed by the ifcensee is developing the I safety evaluations for~the Topics, and csmitments of the licensee's resources to the SEP review and~ implementation. Addressed' shou,1d be the concern whether*other important' a'etivities of the licensee, such as atten-i tion to prevent &tive maintenance and modifications improving plant reli-l ability, are being diverted by SEP. The Report notes that the licensee is planning for an extended' outage in 1983 and for the cutage following com-plation of Cycle XI. In this time scale,' resolutions should be forthcoming from t,ae TMI related items and from the Unresolved Safety Issues. Mcw . should SEP fit into the prioritizing and categorization of significant im-provements in overall safety presumably maridated by the other ongoing NRC programs? Is the scheduling of-SEP improvements comensurate with the appraisals and resources of the regulatoi and the 1teensee? I believe that the SEP integrated managemer.t taan 3hould have a major role in the decision making applicable to SEP-reviewed ~ plants. . Staff Rascense: Thestaffha=andsillcontinuetoencouragethelicenseei.invol.vement-in the.de,elopmenu of prioritized backfitting schedules. However, the SEP . IdagratA Assessment Team does not have the authority to establish imole; mentatian schedules for TMI, USI or generic multf-plant actions 1 nor does the SE'/'. include making a decision regarding the need to implement. these actions Such decisions have.been established outside the scope of SEP in other licensing forums. ~ -7 GPU has indicated that they will submit a proposed pMorit'ized schedule ~ for planned modifications in November 1982,'which will address SEP actions, ~ generic actions and their own planned modifications. These schedules ' ill w be refTected in the final Integrated Assessment report for Oyster. Creek. A supplement to the Integrated Assessment report or in the' 11censee conversion SER, if' applicable, will address the status of. the TMI, USI and ' multi-p.lant actions. 4 (2). Coment: (Isbin) s Appendix F is a very* detailed report on the operating experieness for I the Oyster Creek Nuclear Geesting Station. The review did not uncover'any significant " aging" effec % CnTy two specific conclusions were reached regarding concerns pc "N 'im ':o "... losses of containment integrity..." and ...the outdated or m.. w ', procedures." Perhaps another approach could l be used Which would u mors Atetent and effective.in providing the input needed for.the integrited assessment. Highlights shocM include managment j responses and comitments to NRC. Inspection Reports, independent safety-audits, and industry-sponsored evaluations. These.itas are parth.ily covered I under " Operating and Regulatory Perbr. nance Since January 1,1982." 7 ~ =. ,,m y,,,.,,.m, .-___._.,._.-.,,,.--_._m ,_,.__m..

.. u. _ o 2 1 l ~ Staff Reconse: I Management attention to NRC Inspection Reports, independent safety 1 audits and industry-sponsored evaluations is an area that is. evaluated by the Systematic Assessment of Liceitsen Perfor=ance (SALP) reviews. The j Integrated Assessments Attempt to characterize the up-to-data management performance with the " Operating and IMgulatory Farfomance" section of the j report which is prepared by the appropriate NRC Regional Director. j. (3) Comment: (Ishin) \\ In the conduct of SEP, recognition should be made that there may be modifications and/or improvements which have a streng economic incentive rather than an imediate safety significance, but because of ongoimg NRC burdens and comitments are being postponed.. For example, a loose parts-monitoring system is judged using PRA to be relatively unimportant from ~ the standpoint of reducing risks; however, such a. systen might provide an early warning of loose parts and thereby avoid expensive repairs and down-time. The example is carried further to tilustrate another complicatton. A licensee should be able to use this system with rio cemitments in the Technical Specifications. The present SEP conclusion is that no backfit-ting is required with a caveat "...at this time", since the issue is being considered generically for Revision 1 to Regulatory Guide 1.133. Thus,' the implication is made that decisions ranched now for no backfitting m 5 he changed because of a generic re' solution reached at a. later date. .I de generic resolution utilizes the sarze infomation base a*s that used - in 'the PRA, then obviously the decision making processAs with 'particular emphasis on integrated assessments are being muddled *. To avoid this, as noted previously, the SEP integrated management team should be able to l interact with other NRC groups making decisions 7n datemining what should l be applicable for the SEP-reviewed plants. Staff Resoonse: N As previously noted, the SEP does not inclQde making backfitting decisions on requirements that are bairs implemented generically. Based on the infoma-tion available, the Integrated Assessment Team concluded that backfitting a loose-parts monitoring systen,.in order to confom to current criterta, was - not necessary for the reasons stated in the report...However, the Comittee i I for the Review of Generic Requirements (ChGR) may decide that backfitting l l I such systems to operating reactors should be requir'ed generically, based on " more detailed cost-benefit considerations. ~ j L I l i a....-. .v..

a 3 f. (4) Coment: (Zudans) i Dr. Zudans noted that "the resolution of mest issues is scheduled

  • for a time frame of several years."

1 1 j Staff Resoonse: As noted in Item 1 above, the staff will consider the prioritized modification schedules proposed by the licensee. In addition, the staff noted in the introduction to Chapter 4 of NUREG-0822 that the refueling outages for Oyster Creek are such that the plant will only operate for approximately one year over the next three years. Therefore,. deferral of several actiens to the subsequent refueling outage is considered reasonable. 1 (5) Coment: (Zudans) 4.1 Topics II-3.5. II-3.S.1 and II-3.'C, Hydrology: Dr. Icdans expressed skepticism regarding the licensee's ability to demonstrate a structural capability to withstand groundwater up to 22 ft ms1. and noted the conservatism in that value.. i ~ Staff Resoonse: w-The staff recognizes the ' conservatism in a groundwater ley.;.el, but believes that the licensee can demonstrate such a structural capability (i.e.,;no loss of function under hydrostatic loading). If los's of function is likely,. the licensee may need to establish realist'ic ground t. water levels based upon actual monitoring and decmenstrate no loss of. i function for structures for those levels. The corrective actions do not necessarily' require structhral ugrading. (6) Coment: (Hendrie) l ~ 41 Topics II-3.8, II-3.8.1 and II-3.C. Hydrology: OF. Hendrie noted th t splitting the acergency power sources to the condensate transfer pumps may not be necessary becaose it should be possible to hook up a L portable pump to supply makeup to the isolation condersers. 1 4 t -l e ' ~ ~ I L.

.. - ~. - -

  • e l

4-Staff Resconse: Dr. Hendrie's observation is correct. However, the staff concluded that the cost of pr6viding separated power supplies is justified by the benefit of eliminating the single failure vulnerabiltty, such that theia i . would be an increased capabtitty to provide cooling for flood.and other ' N abnomal event. The staff considers this action preferable; however, the ~' licensee ts currently evaluating thts modtficatten and may propose an alternative. A portable pump is a potehtial alternattve. (7) Coment: (Hendrie) 4.2 Topic III-1. Classification of Structures. S~ystems and Components: Dr. Hendrie did not object to the staff's recomendation, but gu&ssed that i the reexamination of radiographic and fracture toughness requirecents would .i result in some hard questions about present sarviceability. ' Staff Response: The staff's recomendation was not intended to demonstrata conformance with current requirements, rather to establish existing margins of safety on a sampling basts. In the event the licensee cannot demonstrate adequate margins.. then some form of corrective action should be preposed. For. eESle, modification' of component supports to reduce loadings. S (8) Coment: (Bush) 4.7 Topic III-4.B. Turbine Missiles: This problem was no't covered in-Palisades and Ginna and no reason was advanced as to why it is raised With - Oyster Creek. Since this seems -to. be a deviation from the existing nom, it L might merit a sentence as to why it was included; e.g., turbine crientation, more vulnerable safety systams, etc. Staff Resconse: ? The issue is turbine blade and rotor cracks, that could lead to turbtne blade failure thus generating missiles with potentia'l damaging effects to i ESF systems. hurbineMissilesforPalisadesandGinnawerenotaddressedbecausethey use Westinghouse Turbines and Westinghouse has submitted reports for staff I review on establishing inspection frequency. The staff generica11y'specified new inspection requirements for all Westinghouse plants (Palisades and Ginna included) based on their review of the Wertinghouse Reports. , For'.the.GE plants (e.g., Oyster Creek), the review of the infomation regard-~ ir. inspection frequency for GYTurbines has not yet been completed and, therefore, the licensee has been requestad to establish both a basis and insepction frequency schedule. e -e neueue -,y-,- --w.,-,gr--.,g,-,3 ,-.r ,y..,,w~-,-- m--3-y my s -o w,s.m,me, e ew.--w-ww ,-wee y, ..-e--,-m-.e -ee.-w+ r gy.. -y,w,

m..._ ^ s e 5- ~ (.9.1 Comment: (.Hendrie). 4.16 Topic Y-5, Contatment Leakage Detection: Dr. Bendrie questioned the degree of difficulty for qualifying a contatnment leakage detection system to the Safe Shutdown Earthquake (.SSE) level and questioned I the need to require operating limits 1n the technical spec.tftcations for I these leakage detection systems. j Staff Resconse: q Based'.ori'coments received.fran both Dr. Hendrie and the ACRS, the staff has modifted its position to require a reliable leakage' detection system which is sufficiently sensitive to detect the leakage rates predicted for assumed flaw sizes. It is also the staff's position that the assumed flaw size have substantial margin to unstable failure for loading conditions in excess of the design basis. l With regard to the limiting conditions for operation.- Me Technical Specifications, it is the staff's position that operabiltty cd restems i required to maintain the plant within its design envelop should bs contained. in the plant Technical Specifications. Because the. leakage detection systems provide protection against a spectrum of pipe break related events, the, staff concluded that its operability should be included in the Technical Specifica.- tiens. Further,10 CFR 50.36(d)(3) specifies that the Cen:nission may amend a plant license to include Technical Specifications of the scope and content required for new plants. .,;. k .l sl (10) Coment: (Hendrie)

4. 21 Topic VI-1, Organic Material and Post Accident CheImistry:."This issue is new to me.

I understand the intent. but I don't understand the l accident precursors. I wedder i.f,,,tt's significance has been exaggerated." Staff Resconse:- the issue is.to examine the coating daterials in the drywel'1 that could I potentially degrade during an accident condition and impair the function of ESF quipment parts (such as bearing seals or cooling paths) or decompose to 3 . produce hazardous environment (such as hydrogen generation). Also, the torus l i water chemistry should be controlled to reduce the potenttal for stress j corrosion cracking of'senpitive reactor components if the torus water is injected into the reactor. El'iminating a problem source is reasonable when considering the potential adverse effects. = I =e - w ---,.er

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. ~, .g. t i (11) Coment:,(Hendrie) 4.23 Topic VI-7.A.3, ECCS Actuation System: The issue here boils down to matters of testing the emergency cendenser actuation logic provisions and putting the test details in the Technical Specifications. I wculd leave them in the operating procedures. Otherwise the staff proposition is j . appropriate. Staff Resconse: Similar to the response ts Coment 9, the staff concluded that surveillance of safety-related systems', expecially those whose failure could c:mprcmise the i ability to safely shutdown the plant, should be included in the plant Technical j Specifications. t l-(12) Coment: (Hendrie) ti 4.25 Topic VI-7.C.1, Appendix X, Eiectrical Instrumentation and Control TEIC) Re-Reviews: This topic involves automatic bus transfers between i redundant safety-related pcwer sources. It has'long been the conventional wisdom. among staff EI&C. types that autcmatic bus transfers are a "no-no" in essential electrical circuits. There is a reasonable basis: whatever improve-mant.in reliability of supply that results from proper functioning of automatic transfers is more than offset by the chance of switching a fault onto the resaining live bus and thus failing both spurces. It is probably a good rule for new construction. It is not so clear it is a good idea for all of the seven automatic bus transfers the staff wants taken out of Oyster Creek. The seven. loads which have auto transfer between the two emergency AC buses are MCC-IAB2, Vital Lighting Distribution Panel..A. Protection Systems Panels 1 and 2. I don't know what motor control center feeds, but those other panels sound like a lot of instruments and vital lights that one wodid not want to. -lose,'.and. removing the auto..tran(ers leaves them with a single power supply. A PRA estimate was made to see shat improvements in emergency pcwer reliability would result from removal of the auto transfers (Appendix D, page 60). The-result was an improvement of 15%, but is masked by the six times more likely failure of both diesel generators. It was rated as having medium importance., As I read the PRA section, it sisms that some very conservative, assu=ptions were made about' breaker failuret:' Whether the ijnproveent in overall - emergency AC reliability is worth the reducaif reliabiltty of power to all those instruments is a nica question; and I am unimpressed by the evidence in hand. In fact, problems of handling the plant upon loss of the instru-ments powered from Vital AC Panel 1 are looked at in Appendix 0, page 93, in connection with Topic VII-3. The conclusi6n there is that loss of the VACP 1 instruman'ts does degrade operator response significantly, but is masked by other operator error probabilities and so does not contribute to the overall core melt risk. Somehow, that does not give me much ccmfort on the question d taking out these auto transfers. I think the matter deserves seme further. t , analysis. 1:

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[! Staff Resconse: l The intant of the staff position regarding the seven ABT's at Oyster Creek is not necessarily to eliminate AST's but rather to eliminate aut:matic m transf,ars of fault (it is. recognized that' eliminating ABT's will ace:nplish this requirement). The licensee has emmitted to perform a coordinated load and redundant circutt breaker analysis for the subject ABT's to establish a j; design that will prevent transfering of faults onto or between ESF electrical buses. p i ] ~, (13) coment: (Hendrie) ] 4.25 Topic VI-10.A. Testing of Reactor Trip System and Engineered i Safety Features Including Response Time Testing: This topic involves testing 1 of the reactor trtp system and engineered safety feature trtps, including response time tests. The staff conclusions and proposed actions are reasonable, except that I would not put all the-reactor trip system test details in the Technical Specifications. Staff Resoonse: Same as.for Consent 11. y '(147Coment: (Hendrie) 4.36and4.37 Topics XV-ld and XV-18, Radiological Consequences of Small-Line and Main Steam Line Breaks: Mr. Hendrie agrees that reducing the i permissable reactor: coolant activity levels for" lodine (.via Technical Specification 3 changes) is appropriate but questions whether the present react'or water cleanup system is capable of maintaining these low levels without some upgrading. [ l-Staff Resoonse: . Operating experience indicates that BWR plants can and do maintain.the ~ primary coolant activity levels within the limits reccmended by the Standard Technical Specifications. Thus, the existing Reactor Water Cleanup System designs are adequate. 1 g l I-i ? j-l t ~ l

7.- .s.-..__m. ..-.. _.. m._ a. ~# ~ ENCLOSURE 2 OBaielle n u unhws u.%,uoru, B.::cueP. Jm d October 20, 1982 D,'.i'[," I752223 ~ ~l .. w. us j l Mr. Christopher I. Grimes, Project Manager Systematic Evaluation Program Branch Division of Licensing -Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Comission Washington D. C. 20555 F

Dear Chris:

Y Integrated Plant Safety Assessment - Systematic Evaluation.Ctartb /f i

Subject:

8 Procram - Oyster Creek Nuclear Generating Station I Since this is the first of the BWo.'s, I compared the format and general approach to that used with Palisades and Ginna as well as the S.E.P. objectives established by the task force. The formats of all three are quite comparable which I presume is a deliberate decision. The general approach was quite'similar. In fact, I often found . entire paragraphs dealing with explicit. items that were verbatim. 'Comoliance with SEP Objectives I am repeating the objectives and handling of deviations here since this is the first of the BWR's. l 1. The review program must assess the adequacy of the design and l operation of all currently licensed nuclear power plants. 2. The program should establish documentation which shows how each operating plant ecmpares with current criteria on . significant safety issues, and provide a rationale for acceptable departures from these criteria. i ?!~ 3. The program should provide for the capability to make integrated l j and balanced decisions with respect to any required backfitting.

t 4.

The program should be structured for early identification and .j resolution of significant deficiencies. l 5. The program should efficiently utilize available resources and minimize requirements for additional resources by NRC or industry. .i The planned systeutic evaluation wcuid establish the adequacy of all operating' l i power reactors with respect to safety and provide clear written documentation ~l bases for this conclusion. l S2102605c2 821000 CF ADOCK05000$9 gg#7 -~

.= = i s-e @Battelle Mr. C.I. Grimes October 20, 1982 Page - 2 When deviations fmn current licensing criteria are identified, the following alternatives (or combinations of alternatives) will be considered as a basis for establishing acceptability: ~ 1. The deviation can be justified as not significantly decreasing 'the level of safety. 2. Use of non-safety systems to perform safat. ' unctions. 3. Administrative or procedural changes to enhto :s system reliability. d .i 4. Augmented surveillance programs. { 5. Selected backfitting to enhance system reliability. ~, In my estimation the report does meet the objectives. For example, there were several instances of integrated and bala'nced decisions with respect *a backfitting, and I was pleased to note permission to use non-safety systems to perform safety functions rather.than requiring backfitting and upgrading. I believe the report is in compliance with, the five S.E.P. objectives. l 0;eratine Historv I concentrated on the executiv.e sumary to obtain a fiavor of the Oyster Creek operating history and spot sampled the appendix. I feel that the appendix gives an excellent overall picture of Oyster Creeks operation. There appear to be no disturbing trends or major deviations frem the norm. Fornat and Positions Since my original copy had several pages missing, I needed to remedify this section. I touch on items requiring modification below. l e Page xi-The Summary, cites numbers that do not jibe with later sections. For example, 4,3 items are said to meet current l critaria; the asterisked items in section 3, total 40 with the i suspect items being II. 2-A, IX-6, and.XV-8. I can buy the first two being asterisked, but XV-8 is cited specifically in 4.15 as <~ j having been analyzed. I believe this recuires clarification. Page xv.I confess that I am surprised to see the reader referred ~ to the Xalamazoo Public '.ibrary. For Palisades, yes. For Oyster f Creek, that's a long trip. j e A definite plus, is the use and citation of PRA in examining benefit-risk and deciding that backfitting is or is not required. Tne Appendix 0 serves as a basis.- l 4 er e -

i ~ OBatteIIe Mr. C. I. Grimes October 20, 1982 i, Page - 3 i e The consistency of format and of handling the various issues 'i is very important in promoting the feeling that the SEP objectives are being met. The transition from PWR's to a BWR was simplified by this consistency. I compliment the author's. This consistency is prcmoted by the similar handling of i equivalent items in Palisades, Ginna and Oyster Creek. Presumably this will continue through the other SWR's. e ' The turbine missile problem III. 4-8 was not covered in Palisades and Ginna, and no reason was advanced as to why it is. raised with Oyster Creek. Since this seems to be a deviation from the existing nom, it might merit a sentence as to why it was included; e.g., turbine orientation, more t vulnerable safety systems, etc. e An issue I raised with regard to Ginna of not recognizing PRA positions cited in Appendix 0 does not exist here. You are to be complimented on the extensive use and citation of PRA in the decision process. Soecific Conclusioris Rather than presenting a broad overview such as with Ginna, I decided to revert to the Palisades approach of addressing each item separately. This will provide a better basis for reviewing future BWR's. 4.1-(II.38, II.38.1, II.30)' Positive actions as to modifying systems and Tech. Specs consister.t with action in PWR. 4.2-(III.1) Seismic and Quality. Very similar to Palisades. No problems. 4.3-(III.2) Wind and Tornado loading. I applaud the item-by-item analysis permitting a selective approach to evaluate or to accept rather than the l arbitrary "do it all" approach 4.4-(III.3A) Similar to 4.3 in approach - good. 1 r 4.5-(III.3C) Generally in good shape; however, inspection program needs to [ be formalized. l l 4.6-(III.4A) Tornado missile approach also selective as to whether given items need to os considered. Position acceptable. 4:7-(III.48) Turbine missiles. See previous coments. I have no problem with requirements. 4.8-(III.40) Selective handling of road and air traffic. l t l L.-.

.: =. OBattelle Mr. C. I. Grimes October 20, 1982 Page - 4 4.9-(III.5A) I like your flexible position re handling multiple breaks. It would never have occurred a few years ago. Leak detection should help. Incidentally, the PRA position re break / leak time tends to be conservative. 4.10-(III.SB) My biases show; I'm happy to see the Staff accent the low probability of breaks in some locations such as between valve and contain-ment wall. I fcught and lost the battle about ten years ago. 4.11-(III. 6) Sample analysis of seismic loads a good approach. The CRD analysis may be complex. 4.12-(III.78) Your sampling bases on handling of loads is a good approach. [ 4.13-(III.8A) Personally, I feel that the Utilities hurt themselved by not installing loose parts monitors. The recent extensive damage to steam generators requiring a. years outaos mioht have been avoided. I acree this is not a major safety issue most of the time. It can be some of the time. In any event, it can be a major economic issue, but that's not NRC's respon-sibility. 4.14-(III.10A) A combinatier of PRA and selective evalua. tion established that the backfit gain was marginal - a good approach. 4.15-(IV. 2) Similar to 4.14.' FRA and an earlier Dresden-2 analysis served i as basis for decision re acceptability. 4.16-(V. 5.) This is similar to 4.13. Leak detection has both safety and economic implications. Reliable and diverse systems are cheap insurance. 4.17-(V. 6) This tends to be a bookkeeping item. Postion acceptable.

  • 4.18-(V.108)

Again - A gcod us's of PRA. 4.19 - (V.11A) PRA establishes significance and makes this a high priori;.y. A good approach. 4.20-(V.12A) A reasoned approach when the parts do not meet current criteria but th's evi..all program meets intent.

4. 21-(VI.1) This issue is new to me.

I understand'the intent, but I don't understand the accident precursors. I wonder if it's significance has been exaggerated. s 4

~ Mr. C. I. Grimes A. BaMe October 20, 1982 tg Page - 5 4.22-(VI.4) Valves still seem to be the Achilles heel. The PRA serves as 'j a good basis for action.- I concur. I 4~. 23-(VI.7A.3) Testing ECCS logic. Issue resolved. 4.24-(VI.7A.4) OK

4. 25-(VI.7C.1) Selective analysis - no backfit.

tj 4.26-(VI.10A) PRA caused Staff to retreat from earlier position. Fits ( into "Objm:tives" approach. t 4.27-(VII.1A) Upgrading a desirable action. l 4.28-(VII.1B) Another use of PRA. No action although Utility is going ahead. 4.'2g-(VII.2) falective approval. 4.30-(VII.3) ' Use of alternate cool down systems not requiring backfit. Good approach by Utility. Acceotance by Staff equally good. 4.31-(VIII.2) Upgrade on emeriency power diesel - good. 4.32-(VIII.3B) Upgrading of DC System. I agree is necessary. Good 4.33-(VIII.4) PRA establish issues other than penetrations control therefore, no backfit. I concur. 4.34-(IX.5) Selective upgrading of ventilation systems is needed. I 4.35-(XV.1) Use of PRA to establish level of risk (low). Therefore, no ~ backfit. 4.36-(XV.16) PRA. " Ties to 4.37. 4.37-(XV.1S) Decision that keeping RCPB activity limits within GE bounds resolves reitase problem for XV.16 and XV.18. 4.38-(XV.19) Again, a reasoned action not requiring backfitting. It's a good idea to improve MSIV leakage by better maintenance. i-l } 9 'i s ,m, ,w y ,,.,m., ,,7,,--y--,--4, y-9 o --,..pg--p., .-e. ,->_y y ,.-y,e_-... __,,e

..m 1 Mr. C.I. Grimes ?;;:6"e, us2 GBanelle 20 Anoendices A. Duplicates Ginna B. Duplicates Ginna C. Plant Specific D. PRA similar overall to Ginna; of considerable value. E. References F. Operating experience. I concentrated on sumary and skimed remainder. With some cleaning up of items cited earlier, I find this document quite acceptable. Very ly yours, i 961x.x S er H. Bush Senior Staff Consultant SH8/ par G l i G 9 9 l e g t i g ,,y-y.g1-, ._._7w._. ,,y p m, w, my

i ENCLOSURE 2 '81 OOKHAVEN NAllONAL LABORATORY j '{ { { ASSOCIATED UNIVERSITIES, INC. Ucten, t.cng !stend. New Yc.k 11973 I (516)282s g443 Depcriment of Nuciecr Energy FTS 666/ . j Building 197-C October 21, 1982 j Mr. Christopher I. Grimes Project Manager. Systamatic Evaluation t. e Program Branch Mail Stop Slo U. S. Nuclear Regulatory Connission Washington,U. C. 20555 REF: INTEGRATED PLANT SAFETY ASSESSMENT, OYSTER CREEK, ~ SYSTEMATIC EVALUATION PROGRAM

Dear Mr. ' Grimes:

This letter is my technical evaluation rsport on the Oyster Creek Inte-grated Plant Safety Assessment, as given in the' draft report NUREG-0822. It fulfil.ls the. requirements of the first work assignment (Oyster Creek) of Re-vision 1 of 'the project, " Consultant Services 'Dr. J. 51. Hendrie to Review SEP Integrated Flant Safety Assessment Reports," FIN A-3367, B&R No. 20-19-20-21-1. ?- CONCLUSIONS k The Draft Integrated Plant Safety Assessment Report on the Oyster Creek Station supports my previous conclusion, drawn after study of the Palisades and Ginna reports, that the Systematic Evaluation Program is fulfilling the intent of the Consission when it authorized Phase II of the program in 1977. i In reviewing the results'of the staff's integrated assessments, I found two areas where I question the outcomes. The first concerns the removal o.f seven automatic bus transfers betweInn amergency AC power sources (Section 4'.25, Topic VI-7.C.1). I do not necessarily disagree with the ' staff judgments, but question whether the matters at issue have been sufficiently well analyzed to yield the best choice of actions. Details are given in my cements on Section j 4 25. The second area is a general disinclination on my part to see Technical Specifications burdened with details of testing ind surveillance programs. The 8210250220 821021 ADOCK05000g9 g CF ~ +. -,.~u .>,-.._..-.,y,,.--,,,,w--- ,,-.,-.w-, ..--.v..,.-,,,-,v-..-.-- ,-v.

2 specific instances are noted in my detailed ccmments. Aside from these points, I find the staff recommendations for backfitting, or, in a number of casas, that backfitting not be required, to be reasonable and appropriate and the bases for those recommen' atior:s ta be adequate. d As in the Palisades and Ginna evaluations, a number of tr.e staff recom-mandations for Oyster Creek at this stage of the SEP review are for further ^ analysis and evaluation by the licensee. The results of these analyses and evaluations will require a further round of decisions by the staff as to whether or not backfitting of equipment, procedures, and Technical Specifica-tions is required. These further decisions should be made on the same inte-grated assessment basis as those given in the draft report. .l A number of the original.137 SEP safety topics are being treated generic-ally under the Unresolved Safety Issues program, the Three Mile Island Action Plan Program, or other regulatory generic programs such as implementation of Appendix I to 10 CFR 50. These topics have been excluded from the Oyster Creek SEP review reported in NUKEG-0822. ' Resolution of these topics that are specif-ic to Oyster Creek will be needed event. sally. I centinue to believe that the generic resolutions of these topics should be applied to SEP Phase II plants, - such as Oyster Creek, through the SEP Branch integrated assessment' process. Resolutions of the "further evaluation" topics, as well as. resolutions at so'me level of the generic USI, TMI, and Appendix,I, etc. topics will have to be in hand before' any proceeding ~ on the Oyster Creek full term operating license, or the Commission vill. have to exclude them fran the proceeding.. I DISCUSSION i l i THE STAFF SAFETY REVIEW l The intent of the SEP review is to examine a chosen plant against cuIrent licensing criteria and practices in 137 safety topic area's. These 137 topics are listed in Appendix A of the report. Where deviations from current criteria are found, there are a number of alternatives, or combinations of alternatives, j that may be considered as a basis for acceptability. These include acceptance of the deviation t,ecause it does not significantly. decrease the plant safety e -=,ww s., ---,,4..c.,... m .m - - - -,-,.--m.-..,,m,.--,-,n ,,-4...~,_-.,-.- m- ,s-

m ___ i 3-level, use of non-safety-grade systems to perform safety functien's, administ.a-tive c. procedural changes to enhance safety system reliability, augmented sur-veillance programs for the same purpose, and selected backfitting. Ceviations from current criteria are acceptacle if staff evaluations show that the plant 4 would respond satisfactorily to the various' design basis events and that the probability of those or the consequences are not significantly higher than for plants now being licensed in accordance with current criteria. tie Oyster Creek Station has been reviewed against 83 of the original 137 SEP safety topics. The deleted topics, 54 in all, fall,into two categories. The if rst category, which includes 24 topics listed in Appendix B of the re-port, is not covered in the current work because these topics are es~sentia11y the same as TMI Action Plan items or Unresolved Safety Issues that are being treated on a generic basis by other staff groups. The deletion of these 24 l topics from the present Oyster Creek SEP review is reasoaable and is consistent with the deletion of generic items from previous reviews. The resolutions of some,of these TMI and USI items could have major effects on Cyster Creek and ~ its operation. The coordination of these resolutions with the requirements of ~ the SEP review needs carefu1* consideration by the staff. I have reccmmended in 'revious c'eports or. SEP reviews that application of the generic TMI and USI p resolutions, when they are available, should be made to SEP plants by the SEP Branch through the integrated assessment process. I repeat that recomendation for Oyster Creek. The second category of deleted topics, which includes the 30 topics listed - in Appendix C of NUREG-0822, is not covered in the current work because these topics are, for the most part,not applicable to boiling water reactors in gen-eral or to the Oyster Creek site or design in particular. One topic, III-8.S. is covered by a previously published staff report on SWR control red drive l failures. Two cther topics, V-1 and V-2, are concerned with ASME Code confor-mance' requirements that are reviewed under another topic and with possible-future code cases. Two topis, XI-1 and XI-2, concern Appendix I to 10 CFR 50 I matters that aia the subject of generic programs in other staff offices. The Technical Specifications topie., XVI, is to be addressed later, when all of the other SEP topic resolutions are settled. Aside from these six topics, the rest of the 30 topics in this category are of the "not applicable" type. I find these 30 deletions to be proper and appropriate. amm e em o mp y v-r,va--. ,q s,.-- .-w-.9-y,,,ey,.--g=wa grw-=u. w s%-g<m.y..%=.,-.,m.sme-ewn r-rwe4new m-w g-r+m.smw+ei mwam,i-um e - e me -me--ews-e-- --wwe e emev =-ei-rw +- +-i.-c-

r. i J-Of the 83 SEP topics reviewed for Oyster Creek, 43 resulted in staff judgments that the plant meets current criter1a or is acceptable on another de fined basis. Time has not permitted me to sample the staff safety evaluation l l letters on those 43 topics. The remaining 40 topics were subjected to the SEP l i integrated assessment process, to determine what. actions should be taken to ? deal with the identified devi&tions from current criteria. l I judge ~ that none of the 40 topics included deviations regarded by the staff as so egregious as to constitute urgent safty problems requiring ime-l diate action. The draft' report makes no mention of.the point, but does not indicate that any immediate actions are contemplated., I found nothing in the 40 topics that amounted to an urgent safety problem. j My tally of the outcome of the staff's integrated assessment process is as follows. The 40 topics decompose into some 9% subtopics or issues. (I count I six issues under III-1, whereas Table 4.1 lists it as a single issue'. Other-wise, my count' matches Table 4.1.) The results of the integrated assessments for the 92 subtopics were.that: 31 subtopics yielded recommendations for backfitting of equip- ~ ment or procedural and administrat4ve changes; 30 subtopics were found to require no backfitting or other l changes; and 31 subtopics require further analysis and evaluaticn infoma.

  • L l

tion from the licensee. ( A more evenly balanced set of outcomes for "yes", "no", and "maybe" (with regard to backfitting) cenid hardly be imagined. Changes to the Oyster Creek l Technical Specifications are recommended by the staff in six cases among the 92 l subtopics. That is not a large, addition to the license, but it might reason-ably be even smaller. It is 'not necessary to have every' plant specification and trsting requirement listed explicitly in Technical Specifications, which have grown over the years with such details as to be well-nigh unmanageable. The substantial number of subtopics for which the licensee is to de fur-ther analysis and evaluation will give rise in due time to more review work and to another set of staff decisions on what, if any, backfitting should be done. e 1 ..-,,+n--,--,- wD,, ,,...,,------,,-,.-.---.n -m..~ ,--..w. - - -..,.. - ~. -.... -

u For consistency and proper coordination of staff requirements, It is important that.these decisions be made th ough the integrated assessment process, prefer-ably by the same assessment team. CON 1ENTS ON THE INTEGRATED ' ASSESSMENT RESULTS 4.1 Tooics II-3.S, II-3.3.1, and II-3.C: I take it that. reevaluation of hurricane-and rain-induced flood levels results in rather higher levels tharr the original design basis. Nine issues or subtopics have been identified. (1)' Splitting the emergency power sources to the two condensate transfer pumps, to eliminate the vulnerability of both pumps' to a single emer-gency power source failure is a reasonable stap. I assume at least -{ one of the condensate transfer pumps is needed at the new flood levels .because the service water, emergency service water, and diesel-driven fire water r9mps, which could otherwise be used to supply makeup to the isolation condensars, would be flooded otat. I would think that even in flooding conditions, it should be possible to hook up a por-table, pump rig to supply makeup to the isolation condensers. Pe'rhaps that is included in the RHR reliability item (Section 4.18), where up-graded procedures for safe shutdown using alternata sources of makeup water are proposed. (2) I agree that the revised emergency procedure will adequately specify actions in flood conditions and that changes in the Technical Speci-fications are not necessary. (3), (4) The steps proposed are reasonable. (5) This starts out saying a Technical Spe,cification should be developed for safe shutdown and concludes by saying it.isn't warranted. I agree ,with the latter conclusion. Perhaps the first statement should be, "The staff's initial evaluation was that a Technical Specification should be developed....." similar to (2)? (6) The report text puts the PMH flood level at +13.5 ft MSL. An error? It is given as +22 ft MSL on page 4-1. In any event, the proposed procedure upgrading is appropriate. (7),(8),(9) The proposed actions are rearonable. i t

-- - a 'e 1 6-4.2 Tooic III-1: This topic involves seismic and quality classifications of structures, ccmponents, and systems, and an anempt to demonstrate that Oyster Creek. meets current code and regulatory requirements. I suspect it is going to be difficult to prove that a 1967 Cadillac is, in fact', a 1982 Cadillac. The l better question is whether the 1967 Cadillac is serviceable today. The lican-I see has apparently agreed to a wide-ranging reexamination of radiographic and fracture toughness requirements and of pump, valve, storage tank, and piping 3 design bases. I have no objection to the recommendation, although it is a very substantial amount of work, but my guess is that some hard questions about present serviceability will remain when all the results are in. I wound decide those on the basis that for a plant that has operated reasonably successfully for more than a dozen years, a ccaponent is serviceable unless it can clearly _l be shown to be unserviceable, rather than the reverse logic that a component ta ~ unserviceable unless it can be proven to meet current criteria. 4.3 Topic III,2: Wind and tornado loadings fran a site-specific design basis l tornado are evaluated for.various structures. 3everal of the deviations from current criteria initially identified were found by the assessment tea'm to be non-problems. I agree with these judgments and find the actions proposed for the other issues to be reasonable. 4.4 Taoic III-3.A: This topic involves the hydrostatic loadings frem high. groundwater in flood conditions on plant structures. Two of thr'ee initial is-suas are found on reasonable. grounds (no pun intended) - adequate protection in one case, an insignificant long-term groundwater level change in the other -- to not warrant further review. The evaluation agreed to by the licensee in the third case (short-term groundwater levels) is an appropriate action. 4.5 Toofc III-3.C: Water control structure inspections of intake and dis-charge canals, intake trash racks and screens, and rooi' decks, and a formaliza-tion of these inspections are the subjects. Roof deck's are covered by install-ing scuppers, the canals are judged to be adequately covered by present plac-i ticet, and the licensee has agreed to supplement his insp'ection program and l document it. These are appropriate steos and conclusions. The remaining item, I the staff position that intake structure inspecticns should be formalized in [ shift turnover procedures or in. service inspection procedures until intake water level instrumentation is upgraded, is a reasonable acticn. The licensee already carries out these inspections, but has not agreed as yet with the staff position. 1 -,-n-n, n-

~ _.._.._..__ _ _ 4.6 Tecic III-4.A: Tornado missile protection - The Itcensee h'as agreed to undertake evaluations of vulnerability to tornado missiles in several areas. protection will. be provided if found necessary for various components around the mechanical access opening in the reactor building. The licensee argues that the emergency diesels are not necessary~ for safe shutdown because other motive sources for getting water,to the isolation condensers should be avail- ~~ able in a tornado, but will do an evaluation for the diesel generators. I am doubtful that the licensee can make that argument stick, but, like the staff, will await his evaluation. The licensee also argues that tornado missile pro-taction is not no essary for the condensate storage tank, torus water storage tank, and service water and emergency service water pumps. The staff wants protection for sufficient systems and components to, assure safe shutdown via the protected elements. I think the staff is right and has put the requirement correctly - protection is not needed on all these components, but on enough of them to a'ssure safe shutdown. The licensee is correct in saying that loss of all six senice water.and emergency service water pumps, and the diesel-driven - fire pumps'by tornado missiles is a low-probability event, but I find protec ' tion against the spectrum of tornado missiles for a minimum array of safe shutdown equipment hardens the plant against al1 manner of untoward events in - t'hese turbulent times and is a most comfcrting safety measure. 4.7 Tooic III-4.B: Turbine missiles - The inspections and monitoring programs l proposed are reasonable. l l 4.8 Tooic III 4.0: This topic concerns missiles from offsite explosions, air-I - craft, etc. Truck explosions are found to be of little significance, and the licensee will evaluate possible huards from aircraft. The actions are appro-l priate. 4.9, 4.10 Tooics III-5.A, III-5.'B: These topi's involle pipe breaks inside e and ou.tside containment and the possible effects on sa'faty-related ccmponents fran pipe whip, water jets, etc. Four issues or subtopics inside containment were identified and two outside. I find the assessment team's reasoning and judgments to be appropriate, given the regulatory history on these topics. (See also 4.16.) 4.11 Tooic III-6: Seismic design matters - Three of five identified issues yielded requirements for more detailed analyses. These look reasonable to me. e g .,Mg Nr 46 e -,w-a---.- mw.--wwwi.----.--

j . T.. _.m.,..<_ i a When the results are in hand, the staff will need to take a realistic view of l any indicated needs for more pipe supports and restraints. The other two ise j sues, ability of safety-related electrical equipment to function and qualifica-i tion of cable trays, are being dealt with through SEp Plant Owners Group pro- ' grams and the generic effort on USI A-46. That is an appropriate approach for these' items. t 4.12 Tcoic III-7.B: Design codes, criteria, loading combinations, and reactor cavity design - A number of areas of changes in design codes since the Oyster Creek.'esign was completed have been identified, and the staff wants a review l of them for the as-built plant. The licensee believes his seismic design eval- -l cations submitted to date cover the point. Staff admits some overlap, but .l notes varinus areas not previously analyzed. Staff will settle for a licensee review of the staff's work and a sampling basis evaluation of the code changes i as they affect various plant structures. Licensee has agreed and will identify corrective actions if any are indicated by his evaluation. That seems to me a reasonable proposition. 4.13 Tooic III-8.A: This concerns loose parts and core barrel vibration man-itoring. Since backfitting of loose parts monitoring on operating plants is a generic issue in connection with revising Regulatory Guide 1.133, the SEP ques-tion is whether to jump the gun and backfit now. The staff judgment is "no", I and the reasons presented for that judgment,are sensible and ade'quate., 4.14 Tooic III-10.A: This topic involves thermal overload protection for valve motors. Two issues arise -- thermal overload relay setpoints and use of magnetic vs. thermal-magnetic trip breakers. The thermal overload setpoint matter was one of two rated as " medium" in importance in the probabilistic risk assessments done for Oyster Creek and reported in Appendix 0 of NUREG-0822. t The staff wants the licensee to evaluate setpoints for'ESF valve motors and to bypass, the thermal cverload relays where the setpoints cannot be conservatively established. It is a reasonable step, since it will improve the avai7abitity of many safety-related valves. No backfitting was proposed for magnetic trip-only breakers, since the PRA work showe.d negligible effect on overall, risk. I 1 agree with the staff conclusions. (An editorial note: In Appendix 0, page i xvii, the ava11ab'ility advantage of thermal-magnetic over magnetic breakers is l. given as 30'. It should be' 40'.. The Executive: Sumaries in Appendix 0 are, by the way, a most useful addition.), i .m, ,,,-,,,.-~--e....- -m ,.&,w_,,,,a,,u.,, ,.-.,,-,,,-,,.,,.,-,..,n-

~ T -...-._.-~ I -g-4.15 Tooic IV-2: This item, on reactivity control systems, got'into the integrated assessment when the initial review found itself with insufficient information to complete a single failure analysis. Subsequent work by the staff, including material from the Dresden-2. review, showed that Oyster Creek meets ctrrent criteria and the topic.is thus a non-problem. 4.16 Tooic V-5: Leakagedetection-primarysystem-OysterCIeekhasonlysump '~ monitoring of the' three leakage detection systems recommended in Regulatory Guide 1.45. That exacerbates the staff's problems with pipe breaks"in and out ' of containment (see 4.9 and 4.10). The judgments made for those topics that pipes will, leak before breaking a.;e more robust if there are dependable ways to ietact leakage and thus to anticipate and (void high pressure pipe breaks. The j staff therefore wants the full array of three primary system leakage detection l systems, with one of them. qualtfied to the safe shutdown earthquake and detec-tion sansitivity setpoints detemined by analyses under the pipe break topics for leakage amount before break, Technical Specification limits on plant operation when the leakage detection systems are not fully functional, and Technical' Specification surveillance requirements for the systems -- the whole " megillah". - A limited pRA was done for the topic using a small break LOCA. It gave low importance to the leakage detection systems, but is not pertinent to the staff'.s viewpo."..t, which is focused en pipe breaks and effects of pipe whip, jets,etc. The staff position is, helped by the presence at Oyster Creek of two unused leakage detection systems -- for airborne particulates and gaseous radioactivity -- so that curing the condensation problems in these and putting them in service yields the desired three systems. The licensee has i agreed to put the unused systems into operation and to qualify, one of three ( systems for the operating basis earthquake. I will go most of the.way 'with the l' ' staff on this one: the staff's bases for the r,equirements proposed are reason-able. I would want to know how difficult the SSE oualification is, compared to qualif1 cation to the CBE, to be sure that a practicable system is availabl,e, and see if any appropriate operating limits and the surveillance requirements could not go into plant procedure rather than adding to the Technical Specifi-cations. Two 'other issues under this tcpic, interface leakage and coolant inventory balance, were appropriately found not to warrant backfitting. 4.17 Topic V-6: Vesset integrity - Two issues on vessel material surveillance. i capsules and one on the in-service inspection program are reasonably treated i i i

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-pt -- 4 --*,,-g--4y _a-u==k e e r.-* '... .A.- y. o +7 ,-4 .uv. -10 here. The most interesting aspect of this otherwise rather routine topic is the Case of the Missing Capsule -- which ".... is believed to be lying at the bottom of the fuel pool." l l 4.18 Taoic V-10.B: This topic concerns upgrading shutdown peccedures to as-j sure that residual heat ' removal functions can be assured in all circumstances. The staff has correctly' decided not to limit shutdown functions to safety-grade [ systems. I concur in the staff position that the availability of multiple methods of accomplishing the task increases the likelihoco of getting it done. I 4.19 Toofc V-11.A: The issue raised in this topic is protection of the re-actor water cleanup system from the high primary pressure, to avoid an inter-l facing systems LOCA. The current isolation provisions do not meet current criteria, but the licensee claims the.RWC1J system ha.s sufficient relief capac-j ity to withstand a ' failure of the isolation interlocks. The staff wants an analysis showing the relief capacity is sufficient, the consequences of the relief valve sticking open are acceptable, and the' probability of isolation failure is low compared to FSAR-analyzed comparable events. Otherwise,the staff wants redundant isolation interlocks Mr more relief capacity. The staff position is reasonable and its basis is adequate. 4.20 Topic V-12.B: Primary coolant purity specifications are the subject here. The licensee has submitted a revised water quality specification that would meet current criteria. The staff wants the new specifications incorpor-ated in plant' procedures and Technical Specifications. The actions are appropriate. 4.21 Tooic V'I-1: The issues are organic materials that could be a problem in i accident conditions and post-accident control of water chemistry to prevent long-term corrosion-induced failures. The staff proposals are reasonably-based and adequate. 4.22 "Tooic VI-4: Containment isolation - Like most oider plants, Oyster 4 reek l has numerous containment penetrations for which the isolation valving arrange-ments do not fully meet the current rules in this arcane area. The staff has quite properly determined that extensive backfitting should not be required, since the PRA estimates show a low importance to overall plant risk (and since l extensive cutting and rewelding of penetrations and piping is more apt to "~ degrade than improve safety). However, various administrative steps such as i l I ___._.-.__,..,____.__._._____,__,.,__.___.m.,__.__

- -. - ~... ~ ~.. _. _ s -locking some isolation valves closed are recommended. Also, accessibility of the operating stations for certain remote manual valves should ce assured and Twakage detection capability provided for two lines. The staff's bases for the actions proposed and for the no-backfitting decisions are reasonable and ade-quate, and I agree with them. (An editorial' nott: Having two issues under this topic that are both labeled " Remote Manual Valves" is not idcal from the ' ~ ' standpoint of clarity.) t 4.23 Tooic VI-7.A.3: ECCS actuation system - The issue here boils down to ~ matters of testing the emergency condenser actuation ':gic provisions and put. ting the test details in the Technical Specifications. I would leave them in the operating procedures. Otherwise the staff proposition is appropriate. j 4.24 Tonic VI-7.A.4: The issue is core spray effectiveness and arises because i the staff had insufficient iaforms. tion in the initial review stage to complete I its work.. Further infonnation is now in hand and will be evaluated in connec-tion with the review of startup from the Cycle X refueling. A new core spray sparger has been developed and tested and will be installed, on current plans, at the Cycle XI refueling. In any event, the review of this as an SEP issue is decla' red complete, a sensible disposition. 4.25 Topic VI-7.C.1: This topic involves automatic bus transfers between re-dundant safety-related power sources. It has long been the conventional wisdom among staff EIAC types that automatic bus transfers are a "no-no" in essential electrical circuits. There is a reasonable basis: whatever improvement in re-liability of supply that results from proper functioning of automatic transfers is more than offset by the chance of switching a fault onto the remaining live bus and thus failing both sources. It is precably a good rule for new con-struction. It is not so clear it is a good idea for all of the seven automatic bus transfers the staff wants taken out of Oyster Creek. The seven loads which have auto transfer between the two emergency AC buses 'are MCC IAB2, Vital t.ighting Distribution Panel 1, Vital AC Power Panel 1, Continuous Instrument I-Panel 3 Instrument Panel 4, Protection System Panels 1 and 2. I don't know ~ what the motor control center feeds, but those other panels sound like a lot of instruments and vital lights that one would not want to lose, and removing the [ auto transfers leaves them with a single power supply. A PRA estimate was made to see what improvements iri emergency power reliability would result from h N t . _ _. _, _ _._.. _ _. _ _ _ _.. _. _ - _ _ _ _ _ _ ~.. _. _ _ _ ____ _ _

4, 1 3 . l j removal of the auto transfers (Appendix 0, page 60). The result was an im- ' provement of 15%, but is masked by the six times more likely failure of both diesel generators. It was rated as having medium importance. As I read the j PRA section, it seems that some very conservative assumptions were made about ~l breaker failures. Whether the improvectent in overan emergency AC reliability l 1s worth the reduced reliability of power to all those instruments is a nice question, and I am unimpressed by the evidence in hand. In fact, problems of l handling the plant upon loss of the instruments powered from Vital AC Panel 1 l l are looked at in Appendix 0, page 93, in connection with Topic VII-3. The con-clusion there is that loss of the VACP 1 instruments does degrade operator re- { sponse significantly, but is masked by other operator error probabilities and so does not contribute to the overall core melt risk. Somehow, that does not - give me much comfort on the question of taking out those auto transfers..I i think the matter deserves some further analysis. a 4.26 Tooic VI-10.A: This topic involves testing of the reactor trip system and engineered safety feature trips, including response time tests. The staff conclusions and proposed actions are reasonable, except that I would not put ~ all the ieactor trip system test details in the Technical. Specifications. 4.27. Tocic VII-1.A: The issues here concern isolation of flux monitor signals and power supplies for the reactor protection system. The act1'ons proposed are appropriate. ~ 4.28 Tooic VII-1.B: Trip setpoints and drift are the concern in this topic. l The staff recomendations are reasonable and the licensee has decided to put in. an improved trip systen to cure the difficult 1'es. t 4.29 Tooic V'II-2:- The issue here is' resolved by the analysis and. actions of 4.14. 4.30 Topic VII-3: Safe shutdown systems - The staff ' conclusions on water inventory for safe shutdown are correct: depressurization and core' spray from .j the torus will do the job if necessary. Other aspects are covered by the rec-1 ommer.dations of 4.1 and 4.18. . j! ~ 4.31 Tooic VIII-2: Ofesel generators - The issue is loss of emergency power by spurious action by diesel generator protective devices. The staff wants im-provement, in a' nunciation of dissel generator alarms and breassing in accident n I w r**y,e,-, y -. - + - my w--,vvw w y y--- +a.-~.- ,.e ,,_,-m-.-, -.up. w-,, -e-

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i conditions of several protective function relays. A PRA estimate of the worth J of the latter step gave it. low importance, but the staff is properly sensitive to the need for reliable emergency power and recnamends it be done. I agree. 4.32 Tooic VIII-3.8: DC power system monitoring - Like most older plants, Oyster Creek does not have much indication in the Control Room of the condition of the batteries and associated breakers, chargers, etc. The staff wants im-provements -- installation of some battery status alarms. A proper decision on all counts. 4.33 Tonic VIII-4: Electr,1 cal pecetrations - The staff concludes that no backfitting is needed, on reasonable and adequate grounds. 4.34 Toofc IX 4: Ventilation systems - Four ventilation issues appear here: j restoration of usential ventilation units by powering them from the diesel. ge,nerators on loss of offsite powce, and without overloading the diesels; re-actor building ventilation and spread of contamination on its loss; ventilation cooling of core and containment spray pumps; and ventilation and cooling of batteries, motor-gkaerators, and switchgear. The. licensite has concluded the plant can be shut down safely with a loss 'of HVAC. The staff concludes the ~~ reactor building ventilation ' issue requires no 'backfitting, but further evalua-tion by the licenses of the other issues is needed to demonstrate essential components can stand a loss of ventilation and that ventilation loads can be taken by th,e diesel generators, and to identify any needed corrective actions or improvements. The staff position is reasor;able on all counts. ~ 4.35 Tooic XV-1: This topic involves feedwater transients, specifically failure of the feedwater controller. The staff concludes ne actions are needed iri view of the turbine bypass system surveillance program now being carried out, the short time until shutdown for Cycle X refueling (January 1983), and the low importance in a risk sainsa of feedwater contioller malfunction. I agree.' 4.36, 4.37 Topics XV-16 and XV-18: These topics involve. radiological conse-quences of failure of small lines or main steam line outside containment. The i issues in both cases are allowable primary coolant activity, and staff posi-l tions that these accidents should yield conservatively calculated offsite doses i i [ esee.

. -.. ~ 14-well below Part 100 guidelines. Oystar' Creek's current primary coolant activ-1 i ity limits are well above ths SWR standard tech spec values, and doses calcu- ~ lated with the convent 1onal staff models are high. The staff wants the BWR ^ 's l standard. tech spec limits to be used for Oyster Creek and put in the Technical 1 i Specifications. Even then, the small line case would not meet current crite-ria. The staff, for good and adequate reasons., wculd not re.1uire flow limiters in the small lines. I think the staff position is reasonabl'e, but wonder if ^ i the reactor water cleanup system is capable of the job without some upgrading. 4.38 Tooic XY-19: The issue'here is radiological consequences of LOCAs, and -1 the contribution to same from main steam line isolation valve leakage. On rea. I j sonable grounds, the staff does not recommend changss in the Oyster Creek.MSIV i leakage limit or more radical measures -such as a current'model leakage control system. The staff does want an improved preventive mafrtenance program for the ~! MSIVs, or justification that the present program is adequate. Tesi,ingexperi-ence is to be included in the evaluation. The steps proposed are reasonable and appropriate. Sincerely, . A Ye l h M. Hen'drie ~ i f l l JMH/dt e j i I ,- t l L i .ms e e en e =- r-we i.e-- w-w-a Ls

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.m D C.05URE 2 ~ HERGERT O.10. GIN s N sr. LCUts PAVC. MN ss41 g icias saa.un i)- Oc?ober 22, 1982 1 To: Christopher I Grimes, Project Manager ~ Systematic Evaluation Program Branch Division of I.icensing Office of Nuclear Reactor Regslation i Prom: Herb Isbin h Review of 3 raft Ra p, NUREG-0822 IntegraLed Plant safety Assessment Systematic Evaluation P=cgram j Oystar Creek Nuclear Generating Station J ,USNRC, September 1982 y i Previous, Reports: April 23, 1982 Review of craft NUREG-0820 (Palisades Plant) June 15, 1982 Review of Draft NUREG-0821 (Ginna Plant) July 20, 1982 Ccements on the Continuation of SEP, >c-Oyster C:sak is the thi.-d plant to be reviewed in the Systematic Zvaluatieri Progrsa and also is to be considered for the conversion f=cm the Pra 4-4-f cperating License to the Full Term operating License. Completion of the SIP assessment will provida documentation for this step. The SEP review has not i identified any item that was considered to be of sufficient importance to require a prompt resolution. In fact during the course of this review thus far, no modifications and no safety improvements have been implemented,as a result of SZP. The SEP modifications and safety improvements identified can j he implemented on a reasonable time schedule. ~I The SEP reviews start with 137 Topics, for which 30 Topics are not ( applicable for this plant. An additional 24 Topics are relegated for resolu-tion through the, Unresolved Safety Issues and the TMI, Action Plan Items, for which a Supplement to this report will be isstied. Of the 83 remaining Topics, i 43 met c== rent c=iteria or were acceptable on another defined basis. Thus, 40 Topics were considered in an integrated assessment for backfitting. E.w ineering jaWts, licensee's positions, and limited probabilistic risk assessments I (FRA) were used ta foamulate decisions. The following conclusions were I reached: ' 11 Topics (11 issues) require equipent modification or addition 7 Topics (8 issues) require procedure developent or cht.nges j 14 Topics (23 issues) require engineering analysis or continuation of an ongoing evaluation-l 5 Topics do not require any backfitting 8 Issues Licensee does not agree with the NRC position or has not addressed it I. \\ 8210 2so sW.. O- -. 1 4,, .~m- ..,.,_.-,,,.-.,,,s. r., ,-.--m-,e .-w, -,.,-.v ,,m.n, ,.,-.-.-.,-.,._,.ww. -,.,,-,,.,,.,w,,m,,,.,-, c.w,---[..--,

1 t , Cetober 22,'1982 HEREERT 5. ISBIN asis mecanev e.nny. sv. Louis PAAM. MN sset e 4 icias ese. sv i In my judgment, the overall format used to present the issues and decisions is good, as well as the process of decision making. The concise I presentations retain the desired clarity, liut may not reficct the balance in i the decision making as perceived by the licensee. For example, future reperts, 4 and in particular the Supplement; might include interactions of the licensee q._, with industry-sponsored programs as wall with other ongoing NRC activities, the initiative displayed by the licensee in developing the safety evaluations 1 I for the Topics, and commitments of the licensee's resources to the SEP review 1 and implementacion. Addressed should be the concern whether other important -l t activities of the licensee, such as attention to preventative maintenance and modifications improving plant reliability, are being diverted by SIP. The Report notes that the licensee is planning for an extended outage in 1983 and for the outage following completion of Cycle XI., In this time scale, resolu-i tions should he forthcoming from the TMI related items and f=am the Unresolved Safety Issues. How should SEP fit into the prioritizing and categorization i of significant improvements in overall safety presumably mandated by the other ongoing NRC programs? Is the sch dattag of SEP improvements ccamensurate with i the appr=4==? = and resources of the regulator and the licensee? I believe that the SEP integrated maageseht team should have a major role in the decision making applicable to SEP-reviewed plants. y Ih Ltaited probabilistic risk assessaants were made for 19 issues plus a portion of an additional issue. Seventsen issues were considered outside the scope of PRA. Overall, I fcund tho' methods used to be helpful and that the results were utilized in a practical manner in formulating decisions. I App =ad** F is a very detailed report on the operating experiences for the Cyster Creek Nuclear Generating Station. The review did not uncover any I regarding can'cerns per"4a4a*.Caly two specific. conclusions were reachied i significant " aging" effects. to "... losses of containment integrity..." and i i ...the outdated or inadequate pr=cedures." perhaps another approach could he used which would be more efficient and pffective in providing the input needed for the integrated assessment. Highlights'should include management responses and ccmunitments to NRC Inspectica Reports, independent safety' l audits, and industry-sponsored avaluations. These items are partially covered under " Operating and Regulatory Perfo:mance Since January 1,1982". In the conduct of SEP, recognition should be made that there may be modif.ications and/or improvements which have a strong economic incentive rather than an insediate safety significance, but because of ongoing NRC,- burdens and commitments are being postponed. For example, a loose parts e monitoring system is judged using PRA to be relatively unimportant from the 'r standpoint of reducing risks; however, such a system might provide an early warning of loose parts and thereby avoid expensive repairs and downtime. The l i example is carried further to illustrate another cceplicatica. A licensee I should be able to use this system wath no coesiementssin the Technical Specifications. The present SIP conclusien is that no backfitting is required I with a caveat, "...at this time", since the, issue is being considered ^ Thus, the implication { generically for Revision 1 to Regulatory cuide 1.133. l t i l a a -..s

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. ~. - - ..... ~ HERBERT 5. ISEIN amis onysnev mwv. sv. t.ouis panx. wu ss4:e to1as see e4t7 is made that decisions reached now for no backfitting might be changed because of a generic rssolution reached at a later date. If the generic resolution utilizes the same infozmation base as that used in the PRA, then obviously 4 'the decision making processes with particular emphasis on integrated l assessments are bei" 4 muddled. To avoid this, as noted previously, the SEP'

  • integrated meragement teen should be able to interact with other N'ec groups making decisions in dataminhg what should be applicable for the szP-reviewed gau.

of the eight safety improvec:sts required by the staff and to which the-1 licensee has not agreed or has not responded, four relate to Technical speci-fication modifications. It would appear that these items could be readily resolved. Leakage detection is the subject of two items. Local leak detection is to be used as an early " *= for possible breaks in high energy lines in the con *=h==at and as a r=====hle and practical alternative to moasures involving separation and restraints. A difference between the staff's and licensee's positions appears to be confined to be whether the " method" of leek detection is qualified to a safe shttdown earthquake sa h=4e event (staff's position) or to an operating-basis earthquake seisr.ic event Y O I (licensee's position). Further discussions should be held to resolve this M' difference. In another issue, the approach taken by the staff is to require [ a leakage detection capability (with. appropriate procedures) to determine , hen operator actio,n is. required to isolate con ainment spray and core systems. w The licensee has not responded to this requirement. The remaining two safety improvements relate to natural phenomena. One backfit item is to fa = 1 he existing inspection practices and involves the it. ake structure trash racks ) and intake screeas. No changes in Technical specifications ars imposed. The licensee has not responded to this issue. An item of disagreement concer=s tornado missiles and whether redundant, but vulnerable service and er.orgency service water pumps, adjacently located, could be used for accceplishing safe shutdown. In the absence of reasonable estimates for detezzining how unlikely would be the loss of redundarm systems to tornado missiles, I would agree with the staff that any method for accomplishing safe shutdown using systems and r components that are protected from tornado missiles would be acceptable. t

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=. -. -~ a. .,,, a e ENCLOSURE 2 a% .00.Franklin Research Center A Division of The F:enk!!n Inseurs

2. ZUDANS. PN.D.

'l sensemAussaaandcwc, es,cour October 25, 1982 i Mr. C. I. Grimes j Project Manager Systematic Evaluation Program Branch Division of Licensing, NRR U.S'. Nuclear Regulatory Commission } Washington, D.C. 20555

Subject:

Review of SEP Integrated Plant Safety Assessment j Report

Reference:

NUREG-0822, " Integrated Plant Safety Assessment - Systematic Evaluation Program," Oyster Creek Nuclear Generating Stat: ion, Docket No. 50-219 - Draf t Report, September, 1982.

Dear'Mr. Grimes':

In accordance with your request I hava reviewed the cyster Creek SEP Integrated Asse'ssment Report and offer the following comments. i At present, the SEP, program appears to be well organized and well managed. The referenced document summarizing Oyster Creek SEP is comprehensive with respect to the arguments leading to resolution of various SEP. topics for oyster Creek. + 1 The procedure fol' lowed during Cyster Creek SEP review, Figure 1 to 3, is the same as the one used for Palisades and Ginna. As it can be seen from Figures 1 to 3, the procedure is generally well defined and at the completion should lead to 'the satisfaction of the l Commission's goals for the SEP program. Following this procedure, Figure 1, 24' topics exit at (1-) identified as generic items related to USI and TMI Action Plan, 30 topics exist at (2) because these are not applicable to oyster Creek and 83 topics reach (3) where the actual review of SEP topics i for Oyster Creek begins. I iloth & Race Streeu. Philadelphia. Pa.19103 (213) 444 1000 TWX 710 6701889 A_

.. 4.

  • 3 Mr. C. I. Grimes 2-October 25, 1982 USNRC For each of 83 topics, a Safety Evaluation Report was issued documenting the comparison with current licensing criteria and identifying areas of potential backfitting.*

Both Method 1 and Method 2, Figure 2, were utilized in this process. However,*at the stop A (Disporition of Topics), all but 40 were left for back-fit candidacy, the remaining 43 ha~ving been put in one of the i categories 1 to 3. None of the topics fell in the category 4 (i.e., safety significant departure), requiring prompt action. I find that technical arguments leading to distribution of topics j to various categories have sound engineering base. I with respect to forty (40) topics in the group of Integrated Assessment (Sections 4.1 to 4.38), NUREG-0822 represents the Draf t Integrated Assessment Report (DIAR). Resolution of UCI.and TCt 4 topics will be addressed in the final version of NURIG-0822 and a supplement to it. I From the overall point of view NUREG-0822 indicates the same extent. of sound engineering judgement supported by the operating ' experience and limited Probabilistic Risk Assessment as did the reports on Palisades and Ginna. The operating experience (December 23, 1969 to the end of 1981) indicates that equipment f ailures (MSIV, RPC cladding cracks, con-densor tube leakage, isolation ecndensor, isolation valves, vacuum breakers) have been the primary causes of reportable events (644), i human error account for 34% of reportable events. However, major contributor to significant events was the human error (causing 15 of 17 such events). After Region I Systematic Assessment of Licensee Performance (SALP) review, operational and regulatory performance of the Licensee was found to be acceptable with increased attention indicated for areas of maintenance and surveillance. The material presented in NUREG-0822 supports the above Staff's findings. For twenty (20) of thirty eight (38) topics slated fer SIP Integrated Assessment, a limited Probabilistic Risk Assessment (PRA) by Sandia previded logical support. This PRA ranked various SEP topics relative to their importance to the risk. Millstone-1 IREP* system fault trees were modified to represent the failures of the r ~ r l j

  • Reference for correspondence. pertaining to SERs for these topics is in Appendix F of NUREG-0822.

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s Mr. C. I. Grimes Cctcher 25, 1982 USNRC h Cyster Creek systems.* This study identified the tcpic VIII-3.3 l (DC pcwer System Bus voltage Monitoring and Annunicatien) as the { only high risk tcpic since DC power appears in dcminant accident sequences. Many of the topics are rated icw risk (V-11. A, VIII-4, XV-16, XV-18, XV-19) due to the fact that at Cyster Creek overpressure failura of contain=e,nt is virtually assured af ter any ccre melt. The PRA risk ranking of SEP. top'ics, althcugh limited in scope, I appears to be based en a sound icgic. Events dealing with extreme external phencmena, high energy line breaks and changes in contain-i ment st:'ength are not considered in this PRA. i With respect to the specific Integrated Assessment tcpics re-j viewed in NUREG-0822, I offar the fo11cwing additienal ec=ments. Section 4.1 discussas the issues resulting f cm ficed levels calculated by current licensing criteria (II-3.3, II-3.B.1 and II-3.C). Staff's analysis of the impacts associated with these issues is ccmprehensive and the mcdificatiens proposed as a result of SEP review are reasonable.. Os original design of the structures assu=ed a g cund water elevation of 15 f t MSL, the SEP review requires this design to be based en 22 ft MSL. Althcugh the licensee has agreed 'o provide justificatien that the essential t structures can withstand the ' groundwater up to 22 f t MSL (page 4-9, NUREG-0822), I am skeptical about Licensee 's ability to do so. There is scme finite time delay between the PMS (and PMP) and the ground-water level under Cyster Creek structures, hewever, the asswtion that the groundwater level reaches the'PMR during a hurricane is L-very conservative. On Topic III-l (Classifice. tion of StructrIres, Ccmpenents and Systems) adequate information was not available for the SIP. review en radiography requirements, fracture tcughness, valves and storage tanks. Accordingly, the i= pact of current requirements en these items and en the piping systems capability to sustain thermal and cyclic leadings, require further s"W ttals by the licensee. Cn Tepic III-2 (Wind and Tornado Leadings) Staff's findings are that a number of safety-ralated structures do not meet current criteria leads (2.20 mph tornado winds and differential pressures of 1.5 psig). As in the previous topic,. much of the informaticn was not available and many analyses in support of the structural capabilities are in pregtess or are planned. I am in agreement with the S taff's review of this topic. i g This study exa=ined the impact

  • of each topic by qual-tative consideration of Cyster Creek fault trees and by use of insights gained f cm other P'.tAs.

I l e m '*e w-w- -. i-.

Mr. C. I. Grimes 4-Cctober 25, 1982 USh7C ~ With resFeet to the Topic III-4.A (tornado missiles) Staff's position that the licensee previde protection for sufficient systams and components to ensure a safe shutdown is proper in particular since a large number of structures may prove to be inadequace to resist the tornado leads. Staff's findings on Topic III-5.3 (pipe break outside contain-ment) indicate potential for LCCA outside containment due to breaks in main steamline, reactor water cleanup line and isolation condensor steamline (if break devaleps between containment penetratics and the ~ outside isolatien valve, with a single active failure (open) of the inside containment isolation valve). The agreed upon licensee re-sponse'on this matter is appropriate in particular since the arrange-ment of valves is appropriate. Seismic design censiderations, Topic III-6, have been given significant amount of attention and deficiencies, if any, may be found in support structures. I have no problems with the Staff's treatment of this area. In review of Topic V-5 (Reactor Coolant Pressure Soundary Leakage Detection) Staff provides valid arg):ments in need to have a reliable leak detection. Topic V-6 (Reactor Vessel Integrity) reIriew findings indicate lack of concern en licensee's part about the RPV materials surveillance program, Staff's position is e,e:tainly valid. Residual Heat Removal System Reliability (Topic V-11.3) has been given appropriate attention in this review and together with the isolation of high-and low-pressure systems (V-11. A), it represents a key parameter related to reactor safety. Staff's analysis cf Topic IX-5 (ventilatica Systems) has correctly identified that the safety-related pump motors must be qualified for the environment in reactor building resulting f cm loss of ventilation system and that the reliability cf the 3 battery . and motor generator room ventilaticn system must be improved. In general, this reviewer notes that Staff has done very. thorou'gh SEP work on Oyster Creek, has ccme dp with properly justi'fied requirements and has demonstrated sufficient flexibility to accept reasonable arguments proposed by the licensee. It is also noted that the resolution of most issues is scheduled for a time frame of several years, 1 l Very truly yours, j i i fQD,j KL,(f6 l Ianons ::cdans ces s, ~ l encls. -w-wc,---- y x--- e,,, - __-, -,y..,,a e- -,-r -_,n o .mw ,,--ws w ,r-ee wm,,,,,

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