ML20093A548

Statement of Matl Facts as to Which There Is No Genuine Issue Re Issue 6 on Atws.Certificate of Svc Encl

ML20093A548
Person / Time
Site:	Perry
Issue date:	07/06/1984
From:	OHIO CITIZENS FOR RESPONSIBLE ENERGY
To:	Atomic Safety and Licensing Board Panel
Shared Package
ML20093A543	List: ML20093A542 ML20093A548
References
NUDOCS 8407100622
Download: ML20093A548 (14)

Text

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STATEMENT OF bWIERIAL FACIS AS 'IO h'rIIG TIERE EXISTS NO N ISSUE 'IO BE IEARD 4

1. Issue #6 in this proceeding st.ates:

Applicants should install an autcznated standby liquid control system to mitigate the consequences of an anticipated transient without scram.

2. .New 10 CFR 50.62 (c) (4) requires that "(t)he SICS initiation must be automatic and must be designed to perform its function in a reliable manner . . . for plants granted.a construction i

permit prior to July 26, 1984 that have already been designed i

and built to include this feature."

3. Applicants were granted a construction permit on May 3,1977.

4. 'Ihe Perry Nuclear Pcwer Plant is being designed and built such that the SIG will be capable of autanatic initiation. See August 13,

> 1982 letter. frcIn D. Davidson to A. Schwencer (Attachment 2) . .

5. Autanation of the SICS can be achieved at low cost. See Applicants' Supplenental Answer to Sunflower Alliance Interrogatory #22, Second Set, February 29, 1984.

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• l % utwna NthCLEAR REGULATORY incident, manual shutdown was

27. Waconsin Public Serdoe Corpora tion COMMISSION accomplished after 30 seconds, and no [WPSC) 10 CFR Part 50

• core damage or release of radioactwity 26. Pecific Cas and Eectric Company [PCEE ,

occurred. 2Menusee vaiity Authoruy (TVA)

Reduction of Risk from Anticipated On November 24.1981, the 30. Pennsy! ania Power and Ught Cornpany gppat,)

Transients Without Scram (ATWS) Commission invited comments on three Eventa for Ught-Water-Cooied Nuciaar alternative proposed rules relating to 31 Wrginia Doctric and PoWr Company Power Planta (\T.PCO) ..

ATWS {46 FR 57521). Each of the three 32. Arkansas Power med us ht company AGENCY:NuclearRegulatory alternative prop sed rules had the (AP&L1 '- -

Commission. objective of reduction of risk from ..

33 Alabarna Power Company (Alabama) s ACTION:Mnal rule. ATWS and each featured a different st Wiscusin Dectnc Powar cocapuy approach to achieve that objective. One suMuAny:The Commissionis amending alternative (the Staff Rtde) emphasized 35 o Authority of the bbe'ob'ew York its regulations to require improvements (pgggy) .c.m. -

individual reactor evaluation to identify , ,

in the design and operation oflight- needed improvemints. The second so. Yankee Atomic Eectric compeny water-cooled nuclear power plants to (Yank ) L e '. . c alternative (the Hendrie Rule) 37. Public Service Cacpany d hdiana

- redu'ce the likelihood of failure of the emp]d d re1, a reactor protection system to shut,down a' ed cEa$ saNr'hYabi t utilities Service r beny the reactor (scram) following anticipated hardware modifications. The third

. ' (NUSCO) transients and to' mitigate the altemative, proposed by the Utility 39. Carolina Power and Light Company consequences of anticipated transients Group on ATWSin petition for (CP&L). second comment without scram (ATWS) event. The final rulemakmg PRM 50-29. prescribed 40. American Dectric Power Service rule requires the installation of certain specific changes that were keyed to the Corpora tion (received June 24.1962) equipment in nuclear power plants. It type of reactor and its manufacturer.

also encourages the development of a Following are members of the Utility Thirty-nine public comments were Group on ATWS, the petitioner in the reliability assurance program for the received at or close t,o the April 23.1982 PRM-50-29.

reactor trip system on a voluntary basis, deadline for submission of comments.

This will significantly reduce the risk of An additional comment was received on Arkansas Power and Ught Company nuclear power plant operation. Boston Edison Company June 24.1982. Copies of the comments EFFECTIVE DATE: July 28,1984. Connecticut Yankee Power Company may be examined in the Commission's The Detroit Edison Company FOR FURTHER INFORMATioN CONTACTt Public Document Room at 1717 H Street. Morida Power Corporation David W.Pyatt. Office of Nuclear NW., Washington. D.C. The following Culf States Utilities Company Regulatory Research. U.S. Nuclear organizations and individuals provided Maine Yanku Atmic Power Company Regula tory, Commission. Wa shington, nunents Northeast Nuclear Energy Compar,y D.C. 20555 [301) 443-7631, 1. F. L Lewis. Philadelphia. Pennsylvania Pacific ces and Eectric Company ,

(private etizen) Public Service Eectric and Cas Co.

SUPPt.EMENTARY INFORWAT)oN:An Washington Public Pow er Supply System anticipated transient without scram 2. S. I, Hiatt. Mentor. Ohio (privete citizen)

3. Washington Public Power Supply System Baltimore Cas and E!ectric Company (ATWS)is an expected operational Commonwealth Edison Company (wPPSS]

transient (such as a loss of feedwater. Consumers Power Company loss of condenser vacuum, orloss of 4. Standardized Nuclear Unit Power Plant Duke Power Company System (SNUPPS] .

offsite power to the reactor) which is Horida Power and Ught Company *

5. South Carolina Eectric and Gas Company accompanied by a failure of the reactor (South Carolina) Long Island Ushting Company 4 :

trip system (RTS), a part of the 6. General Dectric Company (CE) Nebraska Pubbc Power Dtsmet protection system, to shut down the . Duke Power Company (Dukel Omaha Public Power District reactor.The reactor trip system consists.

a. Alomiclndustrial Forum (AIF) . Pennsylvania Power and Ught CIinparJ ,

of those power sources, sensors. s. Detroit hn (DE) Vermont Yankee Nuclear PowerE. arp.

initiation circuits, logic matrices. m Mississippi Power and Ught Company heaWwn h preference among (MP&L) bypasses. interlocks. racks. panels and commenters for the three alternative

11. Texas Utilities Generating Company proposed rule approaches is as follows:.

control boards. and actuation and (WGCl actuated devices that are required to 12. Camonwealth Edisor. Company Support UtilityRule"(PRhi,50-29) initiate reactor shutdown this includes 13. Combustion Engineering. tncorporateJ wppSS circuit breakers, the controlrods and ,

g, control rod mechanisms. That portion of 14fThe utility Croup " on ATWS. representing 4' 22 utilities Commonwealth Edison the RTS exclusin of the controlrods

15. Combus* ion Engineering Owners Group ' The Utility Group on ATWS and controlrod mechanisms is here 16. Houston Ughting and Power [HL&P) referred to as the scram system. ATWS HLap

17. Portland General Dectric Company accidents are a cause of concem (PCEC) Ebasco because under certain postulated conditions they could lead to severe y,Pbcock and i c x Company B&W) PSEhG FPL core damage and release of 20. Ebasco Services. Incorpora ted ((Eba Gulf scol radioactivity to the emironment. The 21. Pubhc Service Dectric and Cas Company ppgt ATWS question involves safe shutdown (PSE&Cl of the reactor during a transient,if there 22. Carohns power and Ught Company Yadee (Cp&LI. first comment is a failure of the RTS.There have been 23. Stone and Webster Engineering Support "Hendr/e Rule"(Most support precursors to an ATWS; the latest was a . Corporation (Sawl for this option is tentative with many

• failure of the automatic portion of the 2t Horida Power Corporatien (FPL) reservations.)

RTS at the Salem 1 nuclear generating 25. Cutf Ststes Utilities Company (Gulf) u a o ina atation on February 25.1983.In that quneUs mpany quesne

(, ,- Feditat! kuist:r / Vcl. 49. No.124 / Tusaday Juna 20, 1964 / hules and stegul tions 26037 l

i CPAL first comment (muld also be substantial probabilistic risk assessment and long. term shutdown and cooling considered a "No Rde" choice) of the issue for each NSSS sendor, and a would be assured).The Hendrie Rule (4G WPSC value-impact analysis of all three FR 57521), while using much of the same P VEPCO proposed rules.The conclusions are:. information base as the Staff Rule.

j S&W 1.The Staff and Hendrie Rules fail the proposed to resolve ATWS by i i value impact test. establishing a reliability assurance 5 & #-No Ru/# ,

2. Only the Utility Rule is consistent program for systems that prevent or.

SNUPPS with current NRC policies. mitigate AMS accidents and g GE Duke

' SJDe record and notice for the Staff prescribing certain hardware -j and Hendrie Rules are inadequate. modifications which would allow for:(1) i

) AIF in order to resolve the ATWS rule Automatically tripping recirculation g MPal. , issue. It was necessary for the NRCs ' taff pump of a BWR under conditions WGC, .

to evaluate the Utility Group report. . '

CE indicative of an ATWS:(2) . . .. ;

his was done by a techmcal assistance automatica'lly actuating the' standby

, CE Owners' Group contract. ' . liquid control system (SLCS) foi BWRs: }

PCEC A report which provided a critique of CPU

13) providing a reliable scram discharge l the Utility Group comments was volume for BWRs (4) providing for the y B&W 3 prepared by Energy Incorporated PG&E prompt, automatic initiation of the l ., through Sandia National Laboratories auxiliary feedwater system for

) AL ma and may be examined at the conditions indicative of an ATWS: and -

4 Commission's Public Document Room (5) assuring that the instruments .

(P R) at im H Street. Washington, necessary for the diagnosis of and Ind D.C. Also, a summary of 39 public h CP L second comment recovery from ATWS accident' yI mments. as weU at a plan to resolve NUSCO , sequences will not be disabled. Finally.

VU "

American Electric the Utility Rule proposed specific design t h P 1.

The Staff Rule option was favored by m difications for each reactor As proposed in SECY-82-275 and the Ms. S. L Hiatt who comnented that it manufacturer. lt contatned proposals Commission briefing on July 13.1982, a was the rnost stringentof the three that:(a) all Westinghouse reactors have Task Force and Steering Group of NRC proposals, but that it would be better to initiation of the auxiliary feedwater personnel from several offices was system and turbine trip diverse from the return to the implemercation of specific - formed to consider the following P bardware changes thae to require reactor protection system:(b) all alternatives .

evaluation models. Conmenters TVA 1. Promulgation of no ATWS rule or Combustion Engineering and Babcock

. and PASNY stated a peference for including ATWS under the Severe *"U .'llcox reactors have diverse I " Alternative 2A" of Ni' REG-0460'. Vol Accident Program. imtiation of auxdiary feedwater and i 4/which is very similarto the Utility 2. Adoption of the proposed or a turbine trip (similar to Westinghouse) d and a diverse scram system and (c)

Mule. The comments fwm Mr. M 1. Lewis modified version of the Util,ity Group I

did not favor any,of thealternatives. but Rule (pRM-50-29) ex sting boiling water reactors f he pointed out limitatiscs of both NRC- 3. Adoption of the Staff Rule or a manufactured by General Electric have i

proposed rules (limitations of modeling) modification of it; or (1) a means to trip the recirculation

_ and felt that the Commission was not 4. Adoption of those portions of the pumps upon receipt of a signal -

? tully addressing ATW1 Hendrie Rule for which there exists a indicative of an / TWS. (2) a diverse Most of the utility r=menters - - technical basis. - scram system, and (31a modification of the scram discharge volume. Also, new

.f'k a

. ' preferred that the Comnission : neCommission has given careful promulgate no rule on ATWS.Howevef. consideration te all the comments and is (three yean after the rule becomes .

many commenters chose either the now publishing a fini rule.nis final effective) General Electric plants would hj -

Utility Rule or the Herdrie Ruleins the I rule use.,in part l'.e same approach that have a standby liquid control rystem 3 more favorable of the itematives

• is used in the Uthity Group's petition for increased to 80 gpm and all reactor j pre,ented (including sane commenters rulemaking. Prescribed changes, keyed licensees would institute training for within thg Utility GrotF).The No Rule to the reactor's type and manufacturer, operators.

3 Y category d'e scribed absve includes those are set out in the final rule.The costs Basis for Mnal Rule a's Promulgated by who felt that the risksikonrATWS are and values of these changes and of other the Commission already suffiNutly low, plus those who considered changes are discussed in a ke$

recommended c6mbinng the ATWS document on file in the Commission's The vast majority of the commenters l$ rulemaking with otherCommission Public Document' Room, entitled felt that the approach of the Staff Rule 4 activities such as the fevere Accident " Recommendations of the ATWS Task was too open-endedin terms of costs to

'h - Program or the develgment of a Safety Force.- resolve ATWS (e.g..the analyses could Goal' be very costly and time consuming). The

'# The co'mments provided by the Utility Summary of Staff. Hendn.e, and Utility Hendrie Rule was found difficult to Group on ATWS conseted of a three Rules nterpret by most commenters.The volume technical report which includes The Staff Rule (46 FR 57521) would ATWS Steering Group opted to evaluate a review and evaluatian of past NRC have resolved ATWS by establishing generic plants,in a fashion similar to the and industry studies. ageneric but performance criteria (e.g.. there would Utility Group approach, and define the be analyses to verify that Service Level various fixes and estimate the reduction

' A free sinale copy of NtXEC-e460. Vol 4. to the C of the ASME Boiler and Pressure in probability for ATWS sequences as -

estent of supply. may be repeated for pubbc Vessel Code would not be exceeded, each additional requirement was added. l

$ 'sEo[ nNna' NNonor fuel integrity would be maintained, there This would then give a value (reduction l

$ . Document Control UA as Regulatory would be no excessive radioactivity in risk) that could be compared to the  :

7 sx*

7 comminion. wuhinston.rac. sosss. - release, the containment would not fall, impact (cost in dollars) of each l

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26036 Worral A..a / Vol. 4w. No. .n4 / Tuesday, June 2tk 1984 / Rulis cnd higul:tions

. h incremental requirement. There cre 1-IncreasedStandby Liquid Conuel large uncertaint:es in these analyses. System /SLCSJ;f Sac 2/cllfl Automatic Initiation of Standby Uquid b and the detailed results of the analyses Contro! System y i

can be found in the report entitled " .' 'Y# '* I ."I'C'i"8

. One of the alternatives considered by C

.. Recommendations of the ATWS Task borated water into the reactor pnmary the Task Force was an automaticaQy Force"(discussed above). A brief

• ant systemme neutmo absorption C initiated standby liquid control system j ..

discussion of the final rule's provisions, b). the baron causes shutdown of the with a capacity of greater than 86 gpm including value/ impact evaluations,is reacton Ad&on of tMs rystem was i given next: pMposed by the Utility Group on ANS (such as 150-200 gpm). This wov]d have 5 J.

for new plants resulted in a considerable risk reduction ( b Diverse andindependent Auxiliary operating hcens(those e three yearsreceiving after the an(about a factor of seven) after the AR)is . #

EndwaterInitiation and Turbine Trip installed for o ersting plants' > h^

effective date of the final rule).ne 'Unfortunately the cost to do th!s (based 1 forPWRstfSaa2(c)(1) Commission beb, eves that, with the use of the Emergency Procedure Gu!delinea on information supplied by the Utihty His was proposed by the Utility Group an ANS.It consists of Group on ATWS)is on the order of $24 h-,

proposed by the BWR Owners Group equipment 2e trip the turbine and initiate and General Electric that are bemg million per plant and is significantly /

auxiliary feedwaterindependent of the implemented at operating BWRs, impacted by the costs of downtime f.cm i b reactor trip syste:n.lt has the acronym increasing the SLCS capacity for an inadvertent trip which would inject AMSAC, which alands for Auxiliary (or operating plants may insure an intact baron into the reactor water and by the i ATWS) Mitigaticg Systems Actuation containment for isolation trans,ients, costs of downtime for installation in 3 Circuitry. lt has a highly favorable although there is uscertainty in existing plants.The value/ impact does 4 value/ impact for Westinghouse containment failure modes. Because of not favor this alternative for existing h plants U plants' and a marginally favorable the vulnerability of BWR containments gy.

vnlue/ impact for Combustion to ATWS sequences,the Commission ggg ., h cons e o permi afte 4 Engineering and Babcock and Wilcox has determined that this enhanced e ve

) ) ,) , g pitnts. Since it has the potential for a mitigation feature is warranted.The have equipment for automatic trutiation

.. g spurious trip of the reactor which high pressure portien of the ECCS of y

rzduces its value/ impact,it should be BWR/5 and BWR/6 licensees (HPSC)is of the SLCS.Most of those plants designed to minim!ze these trips. injected into spray spargers in the core already have been designed for this 4 a

exit plenum. For these plants, the feature. Also, other plants thst have a Diverse Scmm System:Sa62 (c]/2)and preferred location for theinjection of the been designed and built to include this  :

(cll31 nis was proposed by the Utility borated water from the SLCS is the HPCS line,just extemal to the reactor feature must utilize the feature.The equipment for automatic SLCS actuation j

Group on ANS for General Electric. 5 vesselinstead of the standpipe at the should be designed to performits  !

Combustion Engineering, and Babcock core inlet plenum. A similar location is function in a reliable manner and to .

and Wilcox plants. It has a favorable 3 preferred for those BWR/4 licensees provide high reliability against spurious velue/ impact from the Staffs analysis. with HPCI injection into spargers in the actuation. j;t However, the principal reasons for- core exit plenum. This injection location requiring the feature are to assure emphasis on accident prevention and to provides sigruficant improvement m -

mixing of the borated water, particularly Addicg Extra Safety Valves or Burnable Poisons f 0

obttin the resultant decrease in under low vessel water level conditions One of the alternatives considered by 1 potential common cause failure paths in such as encountered when the EPGs are the Task Force was adding more safety the trip system. It also has the potential 3 followed. This injection location is also valves to plants manufactured by f ra spurious trip of the reactor: .

prefened, since it could prevent local Combustion Engineering (CE) and therefore,it'should be designed to Powerincreases and possih!e power minimize spurious trips. For General Babcock and Wilcox (B&W).his would excursions during the recovery phase of reduce the peak pressure in the reactor Ehetric plants, installation may extend an ATWS when cold unborated ECCS vessel and yield a higher probability of by cne or two days the downtime during water could be added above the core. the plant surviving an ATWS with no a refueling outage. Some BWR/5 and BWR/6 licensees core damage.The peak overpressure A diverse scram system for already have this injection location and could also be reduced by modifying the Westinghouse plants was not a  !

have designed the SLCS accordingly. core behavior (the fraction of the time rectmmendation of the Utility Group on the moderator temperature coefficient is ATWS and was not a clear requirement A y,,,,ff,3,,j,,yf,fj,, pyyp pfpf,, unfav rable)by addicg burnable ef the Staff Rule or the Hendrie Rule. BWRstfSn62/c)/5) poisons.ne Utility Group on ATWS 9 alth:ngh the Utility Group on ANS Recirculation pump trip (RPT) wet estimated that installing larger valve interpreted the Staff Rule to include it. proposed as a rule requirement by tLe capacity could cost up to $10 million per ne system does, however, have a Utility ('.oup on ATWS.This safety plant. A large fraction of this cost is the marginally favorable value/ impact for feature will result in a reduction of downtime for instaUation of the valves. -

Westinghouse plants assures emphasis reactor power from 100 percent to about While the probability of ATWS can be on accident prevention, and results in a 30 percent following a transient (and reduced about a factor of three or more, minimization of the potential for failure to scram) within a minute or so. the value/ impact is unfavorable for this common cause failure pa ths. To assure This proposed requirement has already alternative for existing plants. These full opportunity for public comment. the been implemented on all operational plants au have large dry containments

requirement for a diverse scram system BWRs in response to a show cause and will be most able to mitigate the l fcr Westinghouse plants wiU be order dated February 21,1980.The BWR radiological consequences from an .

! published separately as a proposed rule. owners generaUy agree that this is a ATWS. This rule does not cover j

necessary requirement, and it is being enhanced pgssure relief capacity for -

'ne tutan.uan or e diven. . cram symm included in the final rule far new CE and B&W plants. However, the sigmscancy an ct, th. . he/ imp.cs or mSAC. completeness. .'

CoUunission expects that this issue f I

l .

.t, Federal Regist:r / Vol. 49. No 124 / Tuisday Jun2 26.1984 / Rules cnd Regulatlocis 2S039 I wou!d be addressed during the NRC's 2. A numerical performance standard Guide was des eloped jointly by the q design review of any specific new plant for the RTS chs!!enges and the RTS Commission. the Amencan Nuclear or siendard plant upphcation. unavailability to use as an aid in the Society and the Institute of E:ectnca!

Need for all Control Rods to be Inserted initia and c ntinuing evaluat2on of the and Electronic Engineers.

far PWRs 1 th ' 'F *** Each beensee should establish a goal

., 'd'.9""CF 3 A process of evaluat.ing plant. or benchmark to assess the performance T 4 By using soluble boron for burnup and specific and industry. wide operatin8 of the trip system.The Caramisslan and F 43 xenon control. PWRs normally operate

• st or near 100 percent power with experience to provide-feedback to assess whether the RTS is performing the industry have had considerable disag*eement about the "mrrect" or I

'i control rods nearly out (except for some reliably enough- . " appropriate" value of RTS ~ '

{ , Babcock and Wilcox " rodded" reactors 4. Procedures withm. quafity assurance unavailabihty.11 would be mere fruitful

q,fj which keep one bank inserted for xenon programs to ensure that the RTS for each licusee to have a WW's i

control).Thus. nearly a!! rods are performs satisfactorily in service from a d, 7 available to participate in a scam. reliability perspective.%e frequency of for comparison as the plant operates Q insertion of only about :D percent challenges to the RTS should be as low and generates new data.ne tre'atment -

].h ' " of common cause failures w01 be k

, (:pproximately 20) of the control rods is as practicable.

needed to achieve hot, tero power A pivotal aspect of the ATW.S issue is analyzed in a qualitative fashion to provided that the inserted rods are , the reliabihty of the reactor tnp system determine II there are any significant suitably uniformly distributed. What is (RTS). includmg the control rods, and usly un d e ha lure mo es pre important is the uniformipacing of the the difficulty associated with assessing rods. In installing a diverse scram the impact of co= mon cause failures en by,7arring or using existing owners system, the licensee can allow for the availability of the system to function groups since there is much commonabty partial scram failures if it is when required. All RTS systems are in RTS designs. ..n..,

demonstrated that the rod insertion designed for high availability, yet Each licensee, as part of'the RTS pattern is sufficiently uniformly spaced AM'S precursors at Kah! and Browns unavailability analysis, should examine such that a bot. zero power is achieved. Ferry 3. and the ATWS event at Salem t its maintenance, surveillance, and did occur and were the result of testing requirements. The testing .

Considerations Regarding ReEmbility common cause failures of the RTS.The frequency would be examined to j Assurance , Kahl and Brown Ferry 3. incidents were determine if testingis done too often or As a result of the failure of the Safem desenbed in the Federal Register notice not often enough.The type of testing.

m s Unit I reactor to scram automatical!v on containing the proposed rules which e.g.. completeness and seqoencing of-2 February 25,1983, the NRC conducte'd w as pubhshed on November 24.1981 (46 component verification for operability, en investigation of the events (see FR 57521). The Salem 1 incident would be throughly reviewed. The .

L NUREG-0977. "NRC Fact. finding Task occurred after the proposed rules were nature and frequency of maintenance.

L' Force Report on the ATWS Events at published. . e.g., lubrica tion. cle a ning. calibra tio n. '

i Salem Nuclear Generating Station. Unit An analysis of the RTS should be dimensional verification, physical '

1. on February 25.1983" '). One of the performed using existing methodologies movement, would be reviewed.

- principal findings was the lack of for quantitative evaluation of system Recordkeeping procedures should be

', edequate attention being paid to the rehabnity (c g, unavailability). A fault reviewed. -

reliability of the reactor trip system.The tree and qualitative common cause The Commission bebeves that a .

Salem Generic Issues Task Force failure analys4s should be performed to identify the potentral important faults of reliabihty assurance prop foe the recommended to the Commission that a reactor trip systems should be .

[j '

reliabihty assurance program be the RTS. Examples of, quantitative analysis ihr the RTS are: WASH-1400 developed and implemented.with. clear 1 included in the fmal ATWS rule objective of providing additional ..

?- (NUREG4000. Volume 1. " Generic (the Reactor Safaty Study) *. the tnrhan Point Probabilistic Safety Study '. the assurance that the demred:h!gh r .

Implications of AM'S Events at the ., reliability of the-RTS La indeed achieved L' Salem Nuclear Power Plarit"*). While ZionProbabilistic. Safety Stcdy 8

.and other probabilistic safety studie. and maintsined. Operating experience this rule does not require such a In the United States appears to program the Commission urges the Pe,rfor=ed by Industry at their own trutistive or at the request of the demonstrate. in some instances.that voluntary development of a reliability implementation of AppenducA Co are an e 15-

[, cssurance program for the RTS.

p spg p a

(particularly GeneratDesign. Criterion The reliability assurance program 21) and Appendix B to 10 CFR Part 50. '

E' shouId have the foHowing elements: have been performed or are being and other NRC regulatory requirements

1. An analysis of the challenges to and performed, although some of these ,do ,

may not have yielded the degree of failure modes of the RTS system. "D d tan on , d idance is reliabihty that is possible to achieve considering independent failures - with available technology in a cost.

quantitatively and common cause given in the PRA Procedures Guide, E' REC @m 8. km 1983. 'I'his effective mariner. One reason for this failures quahtatively. An estimate of the failure might be that a reliability cha!!enge rate and the reliability of the . y, standard has not been sufficiently 3,. ,,,,,,,,,io,,,,,,,g, RTS should be a part of the analysis. from uit Dmaion of Tuhmeat infonn.uon and developed not quarititatively set down Document Contro!. U.S. Nuclear Resnalate/y in procedures. Another reason might be

' Copies of Nt' RIG-ort and 1000 may be Coeunissiort W ash ngros DC 205ss P"*h'*'d by camns twin a92-esso or bs nung io a faiIure to understand fully the

• The e am de emanuned at id NRc Pubhe the Pubbcauon Seruces Secuon Docurnent Doewtent Room.1 tr H Street. f&.Washaneton. dominant role pl )ed b) cornmon cause Management BrancA Dmsion of Technicat DE :osss. failures.

Infonneuon and Document Control. U.S Ni.c'est ' Copies of this NURIC may be purchased b)

Regula tory Conunission. % ashansion. D C. to5ss. or calLng (301) 4GM53o cr by wnting to the Resulatory Cornmissiott Washingiort DC :osss; or purchased from the Nonocal Techrucallnformsbon Pubt,ca taan Semces Secuan. Docwnent purcha sed from the Na nonal Technical Informeuen Semce. Department of Commerte. s:as Port Royal Mar.arement Bra nch. Dmaion of Technica' Semce. Deparunent of Commerce, s:as Port Ron al Road. Spnt,gr ietd. V A :nst talormauon and Document Control U.S. Nuclear Road. Spruggfield. VA Zrtet. -

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Vu;. w av. u, i a useau), June 20.1W4 / nuh: siJ hegulations .

i De techniques for a reliabilty requir-d by this rule, the Commission n

assurance program are in existence. is not necessary. it is desirab'e that urm licensees to analyze cha!!enges to sensors in the existing reacter trT {;

They have been applied in an orderly, the plant safety systems. particularly the system not be used to provide the structured fashion in defense and reactor trip system. so as to determined signals for the diverse equipment l;

aerospace applications since at least the where impros ements can be made.

o required by this amendment. Use of the 19003. However, details of its .

(

cpplication to a commercial nuclear Considerations Regarding' System and Equipment Criteria same smor for the existing reactor trip

. system and the &vene eqmpment j-power plant have not been worked out.  ;,

Therefore,it is strongly recommended The Commission place a high " " '

that the development of a voluntary premium en hardware, operating [" 'n ' t s e r s th a*e -

practices and maintenance practices difficult to analyze and which could 1i reliability assurance program, limited to the reactor trip system, be performed which will reduce the frequency of increau h pMcMaj gor coq caun O jointly by the NRC and Industry. challenges to plant safety systems. I'N"# 'N'O8 'Y' " "" N Therefore equipment required by this the sensors for the equipment required eppropriately coordinated with INpO, by this amendment do not have to be rule should be of sufficient quality and g EPRI. and the various owners groups. If reliability so as to perform its intended safety related, there should be this propam is not voluntarily function while at the same time considerable flexibility for using existing implemented in an effective manner, the minimi Ing the potential for transients, sensors without using reactor trip Commission will reconsider the question e.g., inadevertent scrams, which of rdemaking in this area, system sensors. However, there may be p

[h chauenge other safety systems. some cases where the use ofless than The development ofindustry The additional equipment required by safety.related sensors would result m p propams on a voluntary basis has this amend:nent to implement diversity precedence in the evaluation of increased risk from frequent safety y{

)

for auxiliary feedwater system , system challenges or where it would not cperating data for commercial nuclear initia tion. turbine trip, recirculation ,

j be cost effective to use se tsors separate power plants. The industry has pump tnp, and reactor trip, while from those in the existing reactor trip developed the Nuclear Plant Reliability . required to be reliable, will_not have to system.This is partiedarly the case Data (NpRD) System as a voluntary meet au bf Ibrattingetit regturements normaUy applied to safety related- where not using sensors in the existing program for the reporting of reliability , reactor trip system wodd result in the data.De NpRD system is now equipment.The equipment required by H undergoing a program of substantial this amendment is for the purpose of need to install a new sensor connected 0 to the reactor coolant system.This could j e

d e on with reducing the probability of unacceptable

,]'RC eb consequences foDowing anticipated result in significant radiation does to ~ U operational occurrences. Since the personnel making the modifications.

improvement is underway, the NpRD system is a valuable element of a combination of an anticipated Another case would be where -

reliability a ssurance progam. operational occurrence, failure of the instauntion of additional containment existing reactor trip system, and a penetrations would be required. In cases f, ChaUenges to Safety Systems seismic event or an event which results where existing protection system f in significant plant physical damage has sensors are used to provide signals to nis rde concerns itself w: h a low probability, seismic qualification the diverse equipment. particdar [!

mitigating systems which are intended and physical separation criteria need emphasis should be placed on the to reduce the chauenge to plant safety not be applied to the equ:pment required design of Ge method used to isdate de systees due to a low probability ATWS by this rde. In view of the redundancy signal h 6e existing protech '

event. However, the Commission has provided in existing reactor tnp system ta mmimize the potential for

' concluded that a reduction in the systems, the equipment required by this .

adverse electrical interactions.

fr;quency of challenges to plant safety amendment does not have to be systems should be a prime goal of each The equipment required by this redundant within itself.

licensee, and the Commission belie'ves , amendment must be implemented such

~The amendment is to require diversity to those portions of existing reactor tn,p that it does not depade the existing that NIWS risk reductions can also be achieved by reducing the much larger systems, where only mirumal diversity is protection system. This is to be I frequency of transients which call for currently provided. The logic circuits accomplished by making the diverse and actuation devicgs e.g, ctreult equipment electrically independent to the reactor protection system to operate, Challenges to the reactor protection breakers on pressunze(d water reactors) the extent practicable from the existing system may arise from such things as: ' st au ize protection system and by insuring that i d d t g eral the ex}: ting protection system MU I Unreliable components, inadequate post. trip reviews, testing. and tolerance ' continue to meet all applicable safety-

1. ",p po n inj con- on u e fs t i

cfinadequate or degraded control s. related criteria after installation of the Existing reactor trip systems, however, diverse equipment.

systems. Operating experience in Japan indicates a transient frequency that is measure a variety of plant parameters The following table illustrates the substantially less that in the Un: led and utihze a variety of sensor types. system specifications that the staff States.Utihties have categorized Common cause failures in the diverse would find acceptable for the diverse g transients for over ten years but have sensors of existing reactor trip systems scram and mitigating systems.The staff are considered sufficiently unlikely that I not speelficallyinstituted a program to will publish this guidance in a '

additional sensor diversity is not Regulatory Guide or Standard Review reduce them. While not specifically 1 necessary. Even though sensor diversity Plan revision which will also cover [

O l

I

i

) . .

i

.- Fed:r-1 Register / Vol. 49. No.124 / Tudsday. }une 28. 1984 / Rulls and R:gulations 26041 i

=

b

, testing. maintenance, and surveillance. test control. (5) control of measuring and granted an exemption from these Additional!>. the staff willissue eAphcit testinF equipment. (6) inspection. test, amendments if they can demonstrate <

? QA guidance for the non. safety re'ated and operatng status. (7) corrective that ther risk from ATWS is sufhciently cquipment in the form of a genencletter. action. and (8) quahty assurance low. Fa ctors important to th:s {

The generic letter will specify which records. demonstrat'.on eculd be powerlevel. t

, , requirements of the following sections of Exempticos

. unique designfeatures that could f Appendix B are to be applied to non. prevent or rmtgate the consequences of 9

'( Some of the older operating nuclear s fety related equipment:(1) an ATWS. remaining plant lifetime, or $

Instructions procedures and drawings, power plants (e.g. those licensed to remote alting. .

, . (2) document control. (3) inspection. (4) cperate prior to August 22.1969) may be ,

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[rsto-oi) [rs,o.ci)

J.

Ptti ating Syste-s System , at Mulation Pv4 Olverse tratter Trfp aad Auto *atic SLL5 Trip 5ystere actuation for this:

Ana111ery feed.ater . .,g Cul0amCf stGARDING ST5 TIM AND (QUIPottuT SPEC 1f! CATIONS Guidance 4tuetten and Terbine Trip for fwus)*

g Mitfekttne Systems .

System *

(Recirculation Pump "

Olverse peactor

• Trip and Autonatic Stt5 ,

Trip Systese actuation for SWs: Olversity from esisting [quipment df rersity to (Outpaeat diversity to y Aust11ary feedwater Reactor Trip Systen the entent reasonat>1e the estent reasonable ce A tuation and Turbine and practicable to and practicable to 89.

fence s siinf atte the potentf al af nimfre the poteattat **.

Trfy for Pds)*

• for co con cause for coeren cause ,

fattures is reqvf red f ailures is required  %

• sty Deleted Not ewquired but the flot esquf red, but the frcun the sensors from the sensors teptementation siest be taplementation rust be to and f acteding the to, but not incteding, o iE-279) components used to the final actuation such that the esisting 'such that the entsting e' protection systmo protection systre laterrvpt control rod device -e.g.. esisttag e pe=er or vent the scrom circuit breakers say 40 continues to seet all continues to swet all applicable safety applicable safety air header. Circutt tie used for ausillary 7

related criteria. related criterf a.

breakers from dif ferent feed eter initiatten. o manuf acturers alone is not The sensors need not -

endancy . 160t required, not required. Suf ficient to provide the required diversity for be of a diverse design or manufacturer.

U Interruption of control [ststfag protection  %

• rod powr. The sensors systne instruaent. q

.ststing recirculation paso trip eqvf pment Installed in BWits In accordance with need not be of a diverse senting lines may c design or suaufacturer. by used. Se nsors

• evious staf f requirements for the sitigat'en of anticipated transtents without

• am need not be sedified. Entsting protection and InstrWat- ca.

systen instrw.ent. sensing Ifnes should

• sensteg lines r a, tg to selected such that Y esed. Senserf and adverse Interactfees

• f nstrunent s.sens f ag with entsting coatrot e

h Itnes should tie selected systnes are evolded.

' such that adverse interactions with g

esisting. control a, systems are avefded.

Electrical Independence kautred fran sensor boutred from sensor  %

from esisting Peactor output to the final - . output to the finst s.

Trty Systers actuation device et actuation device at E which pof nt non. safety which point ron- a related circuits aust safety related be isolated from safety circutts must be 88 related circut ts. fsolated fran sarety 5, related ctrtofts. y e

C".

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==O- e ipCtapM % ^rms ~ u .-- - Jr- L,-~~--_r s + e. ~

s .

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. .emm . _,

w . un-u (1590-of] ,

(F590-01] .

L pittina'ttna Systems Mitfeattat 51stras pectrculation Fuer $rstas pecirculation rep .

systen Diverse peactor Trip and Autoastic stC5 Diver e Reacter Trtp and Automatic $LCS T Fp Systen actuation for S W s:

Trip 5rstan actuation for OWRs: Aust11ery feed =ater *st Austif ery feedwater

• Actuation and Turbine

+

Actuetton and Turbine Cutosace Guidance .

Trip for Puts)* Trty for Puts)* I g

Not r'equired, unless Inadvertent Actuation The destgm should lie The design should te lRf Phstcal separation from Not reqvf red, enless existing reactor Tegp redundant divisions and redundant divisions such that the frequency such Det the freovrncy g-

- of inadvertent reactor of f aadvertent actu. *-

Systes channels in the esisting and channels In the trips and challenges to actos and challenges reactor trip systen are esisting reactor trip ,

other safety systems to other safety 8 not physically separated, systen are not is statatted. systems fs statatted.  %

The implementation sust. physically separated. ,

be such that separatton The implementation 4 criteria applied to the aust be such that , o

• F entsting protection separation criteria a

. sys en are not violated. applied to the esist- .*

ing protection systuu are cet violated. '

7 o,

tav t rorvaent al For anticipated For entfcipated y operettenal occurrences operettonal occurrences a Qualification only, not for accidents.

only, not for accidents.

Seismic Oualtf tcation Not required. Not required, c 4

Quality Assursace Espitett guldsace Esp 1tett guidance .

cn for Test, Natatenance, =111 be issued in will be issued in en and Surveillance a letter. 4 letter. 5 Safety Related (IE) Not required. but rust Not #9qu1r'ed, but eust be capable of perforintag be capatie of perforetag . @

Poder Supply a en

• safety functions with safety functions with .

pa less of offsite power. ,

loss of offstte power. ,

logic and octuetten device Logic powr rust be

• power sust tie from an from an instrument -

• C tastrument powr supply ~ Peter supply fede- .

E Independent from the pendent from the .  %

powr supp11es for the poner supp1tes for the g esisting reactor trip esisting reactor trip c

system, tsisting ATS

• systes. Eststing RT5 ~

o sensor and lastrument sensor and instrument a channel pe er supplies channel power suopites se may be used provided may be used provided the possibility of the possibfltty of

{

conrion made fatture common wede fativre Is prevented. is prevented. * [

'a Segsfr6d.

Testabiltty at Power pequf red. e O

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• We, pp.1,*,4,, ,

w.

, vue. 2.s. A . . , i A uexisy. June 28. Iw, / L!ce W negulations g

-0 With the promulgati:n cf this final ne Reactor Trip System te one of the power plants and reactors. Penalty.

ADVS rule. the Commission has completed action on PRM-50-29. The most important safety systems at Radia tion protection. Reactor siting }-

a commercial nuclear power plants.

petitioner's ref;uests have been granted However,it is only one of many safety.

criteria, and Report:ng and recordkeeping requirements.

j, in part through the incorporation cf 3J requirements into the final rule which related systems which must be closely Pursuant to the AtomicEnergy Act of (J monitored and carefully maintained to address the fotowing issues:(1)(For GE ensure a plant's safety and reliability.

BWRs)(a) recirculation pump trip 1954, as amended, the Energy It Recrganizahon Act of 19M. as amended.

h

};

following an event indicative of an is my view that a more logical approach ' and sectices 552 and 553 oFTitle 5 of the d to reliability asqurance would be to ADVS. and (b) independent, redundant United States Code, the following .j -

consider such a program embracin8 and diverse electricat initiation of scram those several safety systems which amendment to 10 CFR Part 50 is .M following an event indicative of an experience and analyses show could be ' published as a document subject to O codification.

• ATWS;[2)(For CE and B&W PWRs) significantly improved by such a automaticinitiation of auxiliary program.This program should be $-

feedwaterindependent of the reactor PART 50-DOMESTIC LICENSING OF g, reviewed separately from the ATWS

. protection system; and (3)(For rulemaking effort. PRODUCTION AND UTILIZATION q',

Westinghouse PWRs) automatic FACILITIES '

Furthermere, the Commission should 1s initiation of turbine trip and auxiliary not call upon the industry to implement 1. The a dority citation for Part 50 H feedwaterindependent of the reactor complicated and costly reliability continues to read as foMows: H protection system. The petitioner's assurance programs untilit more request for promulgation of specific Authority: Seca.103.104.141.182.183.186, thoroughly analyzes the concept and t es, es Sta t. 936. e3r. 948. 953. 954. 955. 956. e s provisions within the context of on untilit provides specific guidance. ,

A TWStulemoAing for the following amended, sec. 234. 83 Stat.1244. as amended i }

systems are hereby denied: (1) (For GE aton Anab.s (42 U.S.C. 2133. 2134. 2201. 2232. 2233. 2238 BWRs) a scram discharge volume 2M9. 2282h secs. 201. 202,2% 88 Stat.1142. ]

The Commission has prepared a 1244.124e. as amended (42 U.S.C. 5641. 5842.

}

system [this provision was not included regulatory analysis for this regulation. q 5846) unless otherwise noted.

in the final ATWS rule because The analysis examines the costs and 3 licensees already have installed or are benefits of the rule as considered by the Section 50.7 also issued under Pub. L l installing this system), and (2) (For CE Commission. A copy of the regulatory 95-601, sec.10, 92 Sta t. 2951 (42 U.S.C. ';

and B&W PWRs) an alternate mes.ns to analysis is available for inspection and 5851). Sections 50.57(d). 50.58, 50.91, and shut down the reactor that is diverse copying for a fee at the NRC Public 50.92 also issued under Pub. L 97-415. 96 P from and redundant to the electrical Stat. 2071. 2073 (42 U.S.C. 2133,2239).

portion of the reactor protection system Document Room.1717 H Street. NW~ h Washington.DE Single copies of the Section 50.78 also issued under sec.122. y up to but not including the trip brealers analysis may be obtained from David 68 Stat. 939 (42 U.Sn 2152). Sections y (the final ATWS rule includes a W.Pyatt. Office of Nuclear Regulato y 50.60-50.81 also issued under sec.184. 68 requirement for the installation of an @

Research. U.S. Nuclear Regulatory Stat. 954, as amended (42 U.S.C. 2234). j' alternate shut-down system which must Commission. Washington. D.C. 20555 Sections 50.100-50.102 also issued under include the trip breakers). Telephone (301) 445-7631. sec.186. 68 Stat; 95? 142 U.Sc 2236). .

Additional View of Commissioner Paperwork Reduction Act Statement Fw b ses of 68 Sut. i Asselstine 958, as amended (42 U.SC 2273). I While I approve this rule.1 would his final rule amends information i 150.10 (ak (b). a nd (c). 50.44. 50.46.

have required automation of the collection requirements that are subject 50.48, 50.54, and 50.80(a) are is sued to the Paperwork Reduction Act of 1980 under sec.161b. 68 Stat. 943, as Standby Liquid Contro1 System (SLCS) H4 U.SL 3501 et. seq.).These f:r all boiling water reactors.In amended (42 U.SC 2201(h)): H 50.10(b) -

t addition, while I approve the elements requirements were approved by the and (c) and 50.54 are issued imder sec.

eithe finalrule deating with future OfSce of Managernent and Budget -2611. 68 mai. N9, as amended (42 U.S.C.  ;

approval number 3150.<c11.

reactors. I am not satisffed that 2201(i)): and ii 50.55(e) 50.59(b) 50.70.

sufficient attention has been given to Regulatory Mexibi!!ty Certification 50.71. 50.72. 50.73 and 50.73 are issued future reactors.It appears that under sec.161o. 68 Stat. 950, as #

In accordance with the Regulatory amended (42 USC 2201(o)).

significant additional reductions in the {

Flexibility Act of 1980. 5 U.S.C. 605(b).

ATWS risk caribe achieved without the Commission hereby certifies that the 2. A new i 50.821s added to read as 'j incurnng insurmountable economic follows:

rule will not have a significant economic g costs if such measures are considered impact on a substantial number of small .

during the design phase.1 bebeve this trom anticipated transients wtthout scram entities. This rule affects only beensees rule should not be taken as a barrier to that own and operate nuclear utilization ASWenta for W ws'er cwed ~i further consideration of measures for nuclear power planta.

facilities licensed under sections 103  ?

future reactors that can reduce ATWS and 104 of the Atomic Energy Act of (a) Applicability. The requirements of t risk below that achieved by this rule. 1954, as amended. These licensees do this section apply to all commercial Additional Views of Commissioner not fall within the definition of small light.wa ter. cooled nuclear power plants Roberts bus nesses set forth in section 3 of the (b) Definition. For purposes of this In addition to specifying measures to Small Business Act.15 U.St 632. or section. " Anticipated Transient Without reduce the risk from ATWS events, the within the Small Business Size Scram" (ATWS) maans an anticipated Standards set forth in 13 CFR part 121. operational occurrence as defined in Statement of Considerations which accompanies this rule directs licensees 1.ist of Subjects in 10 CFR Part 50 Appendix A of this par: followed by the .

t:" volunteer" to implement a reliability failure of the reactor trip portion of the Antitrust Classified information. Fire protection system specified in General assurance program for the Reactor Trip prevention. intergovernmental relations.

System. Design Criterion 20 of Appendix A of Incorporation by eTerence. Nuclear this part.

s ,

..l

?y -

Feder:1 R: gist 2r / Vol. 49. No.124 / Tuesday June -26.19M / Rules and Regulations 26045 l

(c) Requiremen:s. (1) Ea ch pressurized non. safety related components each s UPPLEM ENT ARY INFORW ATION: AD E3-water reactor must have equipment from licensee sha!! develop and subrnit to the 14-06. Amendment 3fS87 (42 FR 33245.

sensor output to final actuation device. D: rector of the OfDee of Nuclear Reactor July 21.1983). and te: rgraphic AD TE3-that is diverse from the reactor inp Regulation a preposed schedule for 20-51. issued Septem aer 30.1983. require system. to automatically initiate the meeting the requirements of paragraphs inspection of the win ; outboard flap auxiliary (or emergency) feedwater *

(c)(1) through (c)(5) of this section. Each vane support structu e for cracks. The system and initiate a turbine trip under shallinclude an explanation of the manufacturer has sir ce modified the j yg. conditions indicative of an ATWS.This schedule along with a justification if the outboard flap desigt so that some of the L equipment must be designed to perform schedule cal's for finalimplementation inspections prest. rib -d by these AD's

1. Jts function in a reliable manner and be later than the second refueling outrage are no longer requin d. The repetitive 1 rt independent (from sensor output to the after ju!y 26.1934. or the date of inspection intervals may also be . ,

Qi final actuation device) from the existing luuance of a license authorizing, increased.%e FAA has been advised E reactor trip system. opersiiEabo've 5 percent olfuupower. .

that all airplanes in he world fleet have

,9 (2) Each pressurized water reactor T fiEaEYedule'shelftEe~n be mutually been modified in ac<ordance with the

-s a . manufactured by Combustion ag eed upon by the Commission and manufacturer's Instr actions. The i

Engineering or by Babcock and Wilcox Canadian Departmeat of Transport has t licensee.

must have a diverse scram system from issued an AD which reflects the revised j Dated at Washington.DC this day of19th the sensor output to interruption of day of June 19eA_ repetitive inspection 5.This amendment power to the control rods.This scram For the Nudear Regulatory Commission. incorporates the revised inspections and system must be designed to perform its intervals and supers tdes AD's 83-1446 function in a reliable manner and be Samuel ja ,

. and T83-20-51.

independent from the existing reactor secrewyof the Commu.sion. l trip systemlfrom sensor output to in Dw *Hou M *-N **5 al Dis' airplane modolis manufactured j in Canada and type a ertificated in the interruption of power to the control neo coot rsso ew United States under he provisions cf ds)'Each boiling water reactor must { 21.29 of the Federa: Aviation (3) Regulations and the i pplicab'e have an alternate rod injection (ARI) DEPARTMENT OF TRANSPORTATION airworthiness bilater21 ag eement.

system that is diverse (from the reactor trip system) from sensor output to the . Federal Aviation /tdministration This amendment combines the final actuation device.The ARI system inspection requirempts of two existing rnust have redundant scram air header 14 CFR Part 39 AD's and imposes nq additional exhaust valves.The ARI niust be regulatory or econorric burden on any (Doch et No. 84-NM, 4-AD; Amdt. 39-48821 person. Further,it d(etes inspection designed to perform its function in a reliable manner and be independent requirernents that now are superfluous (from the existing reactor trip system) due to modification d the affected from sensor output to the final actuation el

[C s aircraft, therefore, n(+ ice and pubb,c l

~

device. ,

ActNev: Federal / vistion procedure hereon an unnecessary and l (4) Each boiling water reactor must Administration (FAA). DOT. contrary to the publio interest, and good I have a standby liquid control system ACTION:Fmal rule cause having been siown therefor, the (SLCS) with a minimum flow capacity amendment may be t:ade effective in '

and boron content equivalent in control suuuARY:This da:ument adds a new less than 30 days.

capacity to 86 gallons per minute of 13 airworthiness dire:tive which List of Subjects in 14 CFR Part 39 weight percent sodium pentaborate supersedes two eS: ting airworthiness .

solution.ne SLCS and its injection directives (AD) ap>1icable to the Aviation safety, Aucraft.

location must be designed to perform its Canadair Model C 600 and CL-ool I

• Adoption of the Ame tdment function in a reliable manner.ne SILS. airplanes.nese ATs require repetitive ,

r initiation must be automatic and must inspections of the t utboard flap vane Accordingly, pursu nt to the authority be designed to perform its function in 3

[ reliable manner for plants granted a attachment st uctu e.The manufacturer has modified the o; tboard flaps on all delegated to me by thj Administrator' I 39.13 of Part 39 of the Federal Aviation j construction permit after July 28,1964 airplanes. making same inspection Regulations [14 CFR 3"9.13)it amended 1 and for plants granted a construction requirements unnecessary.Wie rule by addmg the followijg new

[ permit prior to july 28.1984, that have consolidates and terises the inspections airworthmess directive:

a ay de ed and built to contained in the expting AD's. Canadnin Applies to Mohel Cb-ooo-1A11 (5) Each boiling water reactor must

""" N'* (C1,-oool and ModelTL-coo-att (Ct,-

ect) airplanes certificatedin all have equipment to trip the reactor ADDRESSES- The sejvice information categories. Compliance required as coolant recirculating pumps . specified in this ADpnay be obtained upon request to Carradair Ltd. Indicated. g automatically under conditions indicative of an ATWS.This equipment Commercial Aircraf. Technical Services. A. To detect possiblelf atigue c:acks in the must be designed to perform its function Box 6087. Station A Montreal, PQ H3C outboard flap vane support structure, 369. Canada, or ma) be examined at the accomplish the followirig inspections for in a reliable manner. cracks on each side of the aircraft. initia!!y (6)Information sufficient to address shown beldw.

oo ' '

demonstrate to the Commission the FOR FURTHER INFORI AATION CONTACT: wfg d adequacy ofitems in paragraphs (c)(1) Mr. Lester Lipsius. Airframe Section. accomplished, and th ' reafter a t intervals not through (c)(5) of this section shall be ANF-172. New Yor( Aircraft to exceed too hours tibe in service.

• submitted to the Director. Office of Certification Offi FAA. New England 3. y;,ustly inspect.the fonowing parts:

Nuclear Reactor Regulation. Region.181 S. Fra Avenue. Room s. ine flap vane support straps. P/N 600-(d) Implementation. By 180 days after 202. Valley Stream New Yor.k 11581, 10460-U and -23. atJhe inboard and the issuance of the QA guidance for, telephone (516) 79 # -6220. outboard ends of the outbc.ard flap. .

D a .

2 i

= 10

.j'. - -

COMPANY

' g' TH E C L EV E L AN D E LE CT RIC IL!:U MIN ATIN G attuuinAtena stoo . ss Punti souAnt j$' .

ct.tvit AND omo moi e itttanowt irisi e. t.9soo c

e o sex sox =

Serving The Best location in the flotion i

Datwyn R. Davidson August 13, 1982

.cc m son,n ,

sis ttw (NCIN!! RING AND CoNsTRUCitoH .

l Mr. A. Schwencer, Chief ,

Licensing Branch No. 2 -

Division of Licensing .-

U. S. Nuclear Regulatory Commission-

. ~

Washington, D. C. 20553

~

i

Perry Nuclear Power Plant '

' Docket Nos. 50-440;30-M1 .- s

~

ATW5 Mitiration Desien Features ..

Dear Mr. Schwencer ,

between CE! and' members' of NRR, we discussed eur

• In a meeting on July 20, 1982, l, plans for changing several systems associated with the m - -

ATWS event. ~

l current ATW5 rulemaking schedule have made It necessary for us to anticipate poten .

l tial future requirements. We believe it to be in our best interest current design and install these systems during our construction construction scheduleas opposed o r, ,

to until the construction of these systems impacts our The inclusion of these systems is based on the proposed rulemaking and is. '

operations.

l not based'en a belief by. CEI that .these systems are needed to mitigate an AT

! . event. . As such, we maintain our, support of Industry comments on the propos '

l .

l rule. .

t l I The basic changes to be made to the Perry plant include the following: ,

t

(!)

An increased flow cepacity for the Standby Liquid Control System .

l " from 43 gpm to 86 gpm. This will Jnvolve increasing the size of both pumps' suction lines as well as changing the reactor vessel in ~

infio tGCS header. Although tTie design includes n both manua.1 .

nitiation caqa_bility, only_ manual l, ttlation will be -

betttrn~aE The existing pumps wilt be us~ed. ..

I Recirculation Pump Trip initiation (2)

An upgrade to safety grade of the ,

circuitry.

(3)

A control grade f eedwater runback feature. ~

An Alternate Rod Insertion system which is redunda,nt to the Piac (4)

Protection, System scram logic.' jl ffhCkY

? -

A. Schwencer ATWS V.itigation August 13, 1932 Pege 2 Tha details of the above described design wi!! be submitted as an amendmenf '

to the Perry FSAR by 3anuary 1933.

Wo be!!cve that this design along wl'th appropriate entergency chrating procedures cnd training adequately address the ATWS Issue for PNPP. .

Yery trulyYours, l

. ' a, 1

i Dalwy R. davidson Vice President System Engineering and Construction .

DRD:WEC:mb. . .

Jay 5!! berg, Esq.

cc John Stefano

• Max Gildner

• j . .

t se 4

S I

• ' j. l

.i i

g .

e t CERTIFICATE OF SERVICE .*  ! M H" C@J3tmL This is to certi'fy that copies of the foregoing were

• served by  !

deposi.t in the U.S. Mail, first ' class, postage pre 18414fd t.)ig y (s Xsi day of O . /, , ,1984 to those on the

sbrvice list b,elow g 7 g . ,. 3 ..

i i . ,- 00CRDinaf53 * :

. r. s t BRANCH c cr., fc-e Susan L. Hiatt

=-

SERVICE LIST .,,

l t

Peter B. Bloch, Chairman Terry Lodge, Esq.

! Atomic Sa'fety & Licensing Board 618 N. Michigan St.

UlS. Nuclear Regulatory Comm.

1 Suite 105 Washington,'D.C. 20555 .

Toledo, OH 43624 ,

I Dr. JerryzN. . Kline l

• Atomic Safety..& Licensing Board.

• I U.S. Nuclear. Regulatory Commission i-WasEington,'D.C. 20555 I # .

Mr..Glenn O. Bright

• Atomic Safety &. Licensing Board L

! U.S. N'uclear Regulatory Commission. .

Washington, D.C. 20555 ,

Colleen P. Woodhead, Esq.

Office of the' Executive Legal Director U.S. Nuclear Regulatory Commission ..

Washington,.D.C. 20555 ,

I..

Jay.Silberg, Esq. 't

. Shaw, Pittman; Potts, & Trowbridge 1800 M Street, NW . l.

. Washington, D.C. 20036 ,

i;f .

Docketing'& Service Branch - ,

.Offi'ce'of'the Secretary , '

U.S.. Nuclear Regulatory., Commission '

Washington, D.C. 20555 -

Atomic, Safety.&, Licensing. Appeal.Bo'ard Panel ,

U.S. Nuclear. Regulatory Commission - -

. Washington, D.C. 20555  !, . .

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