Text

Partially Deleted Secy Paper,Informing Commission of Denial of Petition Filed W/Director,Nrr Per 10CFR2.206
	ML20034H273
Person / Time
Issue date:	06/17/1981
From:	Malsch M; NRC OFFICE OF THE GENERAL COUNSEL (OGC)
To:	;
Shared Package
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References
	FOIA-92-436 2.206, SECY-81-379, SECY-81-379-01, SECY-81-379-1, NUDOCS 9303160306
	Download: ML20034H273 (4)
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^ADJUDICAkbRY ISSUE l

Unfornaation) i The Commissioners i

F_o_r_:

Martin G. Malsch Deputy General Counsel l

From:

REVIEW OF DIRECTOR'S DENIAL OF 2.2 i

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All Fressurized We.ter Reactors Currently License gl Subject _:

Facility:

to Operate l

To inform the Cctaission of a Denial of a P filed with the Directon. Office of URR, pursuant Furpose_:

to10CFR2.206,which[inouropinion,

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j Review Time June 23, 1981.

l as secre-Expires:

William A. Lochstet,

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On March 29, 1979, N

lear tary of the Environmsntal Coalition on.ucfiled a petition Discussion:

(petitioner),

2.206 with the Director, Office of Nuclear Reac:cr

Pcwer, the l

(the Directer), requesting that (NRC) suspend all Regulation, Nuclear Regulatory Commissionoperating lic (PWR).

In essence, the petition is based on the contention Three Mile j

that the occurrence of the accident at 2 (TMI-2), where some of the fuel was damaged or melted, establishes that safety evalua-l Island Unit kformation in this reccrd wn Sey" h

i ti ns for all operating PWRs are invalid to t e in Ecordance with theJ +

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they are based on defective analyses Act. eumNions

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of the performance of their Emergency Cere Cooling t

extent that FOM-f2-9074 l

The analyses are claimed to be the required calcula-Systems (ECCS).

deficient to the extent that t lated l

tions of cooling performance following pos u loss of coolant accidents (LOCAs) were performed a too limited spectrum of postulated i

LOCAs; and (2) in accordance with what petitioner for: (1) believes must be invalid ECCS evaluation s

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CONTACT:

Juan L. Rodriauez, OGC f

9303160306 921207 63t-3224 PDR FOIA 1

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f:rnance be ca2culated in accordance with' an accept-able evaluation ecdel 2.' and, for a nutter cf pestulated LOCAs ef different sices, locatiens.and!

other prcperties sufficient to provide assurance that the entire spectrun tf postulated LOCAs is covered.

On May 29, 1981, the Director denied the instant 2.206 petition.

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inadequacy of the analyses consists cf the foll: wing.

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u March 23, 1979 Three.'.ile Islanc accicent by the Office cf inspectior and Enforce =er.

concludec tna the various accicent and tran-sient analyses performed prier te the accident with respect to the expected performance of TM1's ECCS, were adequate to-have prevented the sericus consequences of the accident, if-the ECCS had been pernitted te functicn as designed.

I' That analyses perfcrned by Eatelle Columbus

.. _ Laboratories (!!UREG/CR12.9." Analysis of the Three Mi'.e Island Accident and Alternative 2ecuences") and by Eateock and Wilecx, ("EJal-uation of Transient Eehavior and Stall Reae:cr Cociant System Ereaks in :ne 177.uel Assently Plant") confirmed that the TMI ECCS system was designed te cepe with the events which ' initiate:

the TMI accident and would have successfully coped with accident had the system not been overriden by the operators..

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That the continued safe operation of the PERs and the adequacy of the ECCS design were confirmed by the report of the NRR Bulletins and Orders Task Force (B&OTF), (NUREG-0645

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Included in this set of criteria, is a requirement that.the maximun fuel element cladding temperature shall not exceed 2200' F.-

(10 CFR 50.46(b)(1).)

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Appendix K to the Commission regulations sets forth certain require:

features of acceptable evaluation models.

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" Report of the Eu11etins and Orders Task Force

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the Director ncte.i..that the.

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_n this regard 3&CTF generically evaluated feedwater tran-sients, small break LOCAs and other TMI-2 1

related events to establish.the basis for t

their continued cperation.

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That as a result Of the TMI Action Plan Requirements (NUREG-0660, Task.II.E.2), small j

break LOCAs evaluatien models are being reas-l sessed with a view to improvement.

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~ 3efore The :!.: clear Regulatory Commi.ssion j

Emergency Petition' f,

The Nuclear Regulatory Conmission is hereby petitioned to j

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irc. mediately suspend the operating license for ea ch Pressurized

'r.'ater Rea ctor (Ph'R) currently licensed te -operate. This petition is broug.ht before the Commission pursuant to j

'10CFR 2.204,2.206, 50.54 and 50.100. ' This petition ~is - brought because the safety evaluations performed for these licenses are

.not valid.

j; As required by 10CfR 50.L6,. ea ch Pt?.. shall be equipped with an Dnergency Core Cooling System (ECCS) uhich shall be designed f

so that its.nerformance shall be to rescond to.costulated accidents in such a way as to limit the fuel temperature as:

I specified in 10CFR 50.L6. ~ This specification would not permit the fuel to melt. These postulated a ccidents include the

- double-ended rupture of the largest pipe in the reactor.

coolant system. All

?'..'?. power plants are licensed ' subject' to this a

regulation. The crocedures.used to demonstrate this ' performanee; o

t are not valid.

y The Three Mile Island Unit II plant is licensed -to operate I

subject to this reoudrement. On 28 Earch 1979 an accident occured -

l at this plant. The accident was initiated by an event less1 l

severe than the double-ended rupture of. the.lar&est ' pipe in the reactor coolant system. The consequence of this accident.

was that at least some of the fuel melted. This consequence is clearly in excess of the performance recuired under 100FR '50.46.- Thus, the analyses used to predict such performance.

i are invalid. Thus, it.has now been established in fact' by.the l

evidence of 28 March 1979 at Three Mile Island Unit II, that-the basis for granting. these licenses was invalid, and they i

should be ir. mediately suspended, or revoked.

-Wh)& h W. A. Lochstet, Secretary

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29 F. arch 1979 j

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MAY 2 9 1931 09 91.D o,,

t Mr. '.fillice Lochstet, Secretary Cnvironnental Coalition on Muc1 car Power 119 E. Aaron nrive State Collene, Pennsylvania 16P01

Dear Mr. Lochstet:

This letter is in response tn your petition dated f4 arch 29,1979, on behalf of the Environnental Coalition on !!uclear Power, requesting that the Nuclear Reculatory Comission suspend the operating license for each pressurized water reactor. The petition alleges that the safety evaluations perforraed for these licenses are invalid because the incident which occurred on March 28,1979, at Three liile Island Unit 2, a pressurized water reactor, shows that the evaluations of the Emercency Core Cooling Systems (ECCS) for all pressurized water reactors cannot demonstrate the performance of the ECCS required by 10 CFR Section 50.46.

I regret the delay in responding to your petition. Our consideration of your request took longer than it normally would have because of the heavy workload of TMI-related natters and the disruption of our routine-response and tract".ing processes which resulted from the T111 accident.

Your recuest has been considered under the provisions of 10 CFR Section 2.205 of the Comission's regulations. This office has determined, for the reasons set forth in the enclosed " Director's Decision linder in CFR 2.205*, not to issue orders suspending operating licenses for each pressurized water reactor.

A copy of the decision is beino filed with the Secretary for the Con,ission's review in accordance with 10 CFR Section 2.206(c). As provided in 10 CFR 2.205(c), the decision will constitute the final action of the Comnission tventy-five (25) days after the date of issuance of the decision unless the Comission, on its own notion, institutes a review of the decision within that time.

A copy of the !!otice of ~ Issuance of the Director's Decision, which is being filed with the Office of the Federal Register for publication, is also enclosed.

TECHNICAL PORTION ORIGINATED BY MARVIN MENDONCA AND NORMAN LAUBEN -

Sincerely, PARAGRAPH 2 ADDED BY RALPH CARUSO f '. ' -

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- SEE ATTACHED YELLOW __..

R.R.Buton

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NUCLEAR REGULATORY CO'illSSION OFFICE OF NUCLEAR REACTOR REGULATION HAROLD R. DENTON,. DIRECTOR In the Matter of

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FETITION TO SUSPEND ALL

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Emergency Core Cooling Systems OPERATING' LICENSES FOR

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PRESSURIZED WATER REACTORS

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DIRECTOR'S DECISION UNDER 10 C.F.R. 2.206 By petition dated March 29, 1979 the Environmental Coalition on Nuclear Power (ECNP) requested that the Nuclear Regulatcry Comission (NRC)- suspend _ all operating licenses for pressurized water reacters (PWRs). This petition has been considered under the provisions of 10 C,F.R. 2.206 of the Commission's regulations. Notice of receipt of the petition was published in the Federal Recister December 6,1979 (44 FR 70241).

The petition contends that safety evaluadons'for all operating PWRs are invalid and thus licenses for all PWRs should be suspended or revoked. Petitioner asserts that tFe consequences of the accident _ at Three Mile Island Unit-2 (TMI-2), at least some of the fuel melted, was in excess of the perforrance required for the Emergency Core Cooling System (ECCS) under 10 C.F.R. E0.46.

Yet, the petitioner also contends, the accident which initiated the.TMI-2 fuel da:nage has less severe than accidents specifically analyzed to demonstrate

-acceptable perforrance by the ECCS. Thus, petitioner contends, the aralyses used to predict perforrance under the provisions of 10 C.F.R. 50.46 must be invalid and hence the basis for granting all PWRs licenses is invalid and these licenses should be suspended or revoked.

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t I have reviewed the information submitted by the Environmental Coalition on Nuclear Pcwer and the issues addressed in the petition.

For the reasons set forth below, petitioner's request that all operating licenses for PWRs should be suspended or revoked is denied.

i Section 50.46 of the Commission's regulations requires that each boiling i

and pressurized light water nuclear power reactor must be provided with an emergency core cooling system designed in such a w

that its calculated cool-ing performance following postulated loss-of-coolant accidents conforms to a set of criteria.

Included in that set of criteria [10 C.F.R. 50.46(b)] is a recuirement that the calculated maximum fuel element cladding temperature shall not exceed 22000F 10 C.F.R. 50.46 further requires that ECCS cooling performance is to be calculated:

1) in accordance with an acceptable evaluation model j

and 2) for a number of post'ulated loss-of-coolant accidents sufficient to provide assurance that the entire spectrum of loss-of-coolant accidents is The spectrum of acciden*s examined includes a break equivalent' in covered.

y size to the dcuble-ended rupture of the largest pipe of the reactor coolant

-l system.- (10 C.F.R. Part 50, Appendix K, I.C.1.)

On Parch 28,1979, TMI-2 experienced a feedwater transient that, through a particular sequence of failures, led to a small break loss-of-coolant accident and resulted in significant core damage.

The failures that were experienced occurred in the general areas of design, equipment malfunction, and human performance.

This TMI-2 sequence of events and failures had not been previously aralyzed and the fuel damage was beyond that predicted by 10 C.F.R. 50.46 analyses. _Therefore, l

a question could be raised as to whether the analyses perfonned to meet 10 C.F.R..

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, l 50.46 were adequate, specifically: 1) whether the evaluation model used for compliance with 50.46 to evaluate the behavior of the reactor system during a postulated loss-of-coolant accident was adequate; and 2) whether there is sufficient assurance that a proper set of loss-of-coolant accidents has been i

analyzed to determine that the ECCS will perform as required.

In the NRC's Office of Inspection and Enforcement investigation of the TMI-2 accident (NUREG-0600, " Investigation into the March 28, 1979 Three Mile Island Accident by the Office of Inspection and Enforcement"), it was stated that the i

TMI-2 accident could have been prevented in spite of any known or postulated inadequacies in tran2ient and accident analyses. The forward to NUREG-C600 states:

"Ihe design of the plant, the equipment that was installed, the i

various accident and transient analyses, and the emergency procedures i

were adequate to have. prevented-the serious consequences of the acci-dent, if they had been permitted to function or be carried out as planned.

For example, had the operators allo ed the emergency core cooling system to perfore its intended function, damage to the core t

would most likely have been prevented."

NUREG-0600 estimates that during the. initial 3h hours of the accident the average ECCS flow was only about 25 gpm, because the operators had reduced the'

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As part of the TMI Inquiry, Battelle Columbus Laboratories explored flow.

i alternative accident sequences.

(NUREG/CR-1219, " Analysis of the Three Mile Island Accident and Alternative Sequences"). They concluded that if the high pressure injection (HPI) ECCS flow had not been throttled, full ECCS. flow l

through the pumps would have remained above 800 ~gpm.

As,a result, the core i

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I would have remained covered, the fuel cladding temperature would not.have in-creased at all, and-the requirements of 10 CFR 50.46 would not have been 1

violated.

Babcock and Wilcox has analyzed small break accidents similar in size to the TMI stuck open PORV.

For these analyses (" Evaluation of Transient Behavior-and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant", May 7,1979) B&W used methods which comply with the requirements of 50.46 and Appendix Kto 10CFR50.. The required single failure for these small breaks would mean that one of the HPI pu ps did not work. The B&W analyses shewed i

that the core remained covered for these small breaks even with half the total

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prssible HPI flow.

Both the BMI and B&W analyses were benchmarked successfully against the reduced HPI flow data from the TMI accident.

Clearly the TMI ECCS system was designed to cope with a TMI type accident.

I and would have,had the system not been overridden by the operators.

10 CFR 50.46

-[

and Appendix K require that _the ECCS be capable of mitigating the effects of an f..

accident, assuming the most limiting single f ailure.

However, it is reccgnized that the occurrence of multiple equipment failures and/or operator errors could result in conditions dich exceed the core thermal limits of Appendix K.

Such was the case at TMI.

Contrary to the petioners contention, the events which occurred at TMI removed the plant from its design envelope, and placed it in a more severe condition than that required to be analyzed by Appendix K and 50.46..

It is not reasonable to require protection from the effects of every conceivable combination of errors which could occur, without limiting the number of errors, because the number of such combinations is limitless.

One of the tasks of the NRR Bulletins and Order Task Force (B&OTF) fomed in May 1979 was to generically evaluate feedwater transients, small breaks 1.0CAs, and other THI-2 related events in operating plants to confirm or establish the f

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-S-basis for their continued safe operation.

In order to fulfill this charter, B&OTF investigated a large spectrum of small breaks and transients to assure that the installed system for all modern operating light water reactors could adequately cope with these events.

Reactor vendors, NRC consultants, and the NRC staff were required to analyze hundreds of cases in pursuit of this goal.

As a result of this review, some parts cf the analytical models were targeted for future review and possible improvement. However none of these sub-models were judged to have a substantialimp6ct on ECCS system design.

The analysis also aided in assessing operator guidelines for recognition and mitigation of small break LOCA (see NUREG-0645, " Report of the Bulletins and Orders Task Force" January 1980).

Another major charter of B&OTF was to assure that all licensees were well trained in the recognition and mitigation of small break LOCAs and would not prematurely throttle or terminate the ECCS during such an event.

To implement therecomendations of all the internal and external TMI inquiries,a planof action was devised (NUREG-0660, "NRC Action Plan Developed as a Result of the TMI-2 Accident," U.S. Nuclear Regulatory Comission, pub-

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lished May 1980 and revised August 1980) in the form of a set of findings and requirements for safe operation of all reactors. These findings and require-ments have been further elaborated in NUREG-0737, " Clarification of TMI Action Plan Requirements,"' U.S. Nuclear Regulatory Comission, November 1980.

In-cluded in this action plan is the small break model re-assessment. This activity-has already begun and includes the comparison of affected analytical model,s to a large variety of experiments.

To date the ability of these analysis tools

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has been encouraging.

I see no evidence that they are not up to the task.'

The NRR staff constantly encourages small break evaluation model holders to make t

improvements.

t In view of the above actions taken since the TMI-2 accident,-I find no basis 4

our to conclude either frem the assertions in the subject petition or from current knowledge of loss-of-coolant accident analysis methods, that the analyses performed in compliance with 10 C.F.R. 50.46 are not valid.

In adoition, as part of.the TMI-Action Plan Requirements, a program to evaluate the uncertainties which may exist in small-breat ECCS performance calculations has been proposed. Holders of approved ECCS evaluation models will evaluate these uncertainties; the Office of Nuclear Reactor Regulation will evaluate their results.

If changes are needed in the analysis methods to properly account for these uncertainties, recomendations will be made to the Commission i

to adopt such changes. (See NUREG-0660, Task II.E.2)

On the basis of my conclusion that the analyses performed in compliance with 10 C.F.R. 50.46 are valid and in view of the many changes which have been j

imposed on pWRs, I find that continued cperation of PWR's poses no undue risk to

.the public health and safety.

[

CC'JCLUSION Based on the foregoing discussion and the provisions of 10 C.F.R. 2.206, f

I have determined that there are no adequate bases for suspension of PWR operating licenses. The request by the Environmental Coalition on Nuclear Power is, therefore, denied.

A copy of this decision will be placed in the Corxnission's Public Document i

Room at 1717 H Street, N. W., Washington, D. C. 20555 and each local public document room for all PWRs.

A copy of this decision will also be filed with the Secretary for the Cornission's review in accordance with 10 C.F.R. 2.206(c),

of the Comission's regulations.

l i

7 As provided in 10 C.F.R. 2.206(c), this decision will constitute the final action of the Comission twenty-five (25) days after the date of issuance, unless the Co=nission, on its own rnotion, institutes a review of this decision within that time.

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Farold R. Denton, Director Office of Nuciear Reactor Reguiation Dated at Bethesda, Maryiand this 29th day of May,1981.

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Task 21.E.2 May 1980 t

TASK II.E.2 EMERGENCY CORE COOLING SYSTEM A.

OBJECTIVE:

Decrease reliance on the emergency core cooling system (ECCS) l for other than loss-of-coolant accidents; ensure that the ECCS design-basis'

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reliability and performance are consistent with operational experience; reach o

better technical understanding of ECCS performance;-and ensure that the uncertainties associated with the prediction of ECCS performance are properly treated in small-break evaluations.

B.

NRC ACTIONS P

1.

Reliance on ECCS.

a.

==

Description:==

NRR will instruct all operating reactor licensees to provide a report that details experience with ECCS actuation (conditions, cause, f requency, results, etc.), compares cumulative experience with design bases for ECCS, and assesses the reliability of the system to perform its intended function under these conditions.

Subject:

" Studies to Improve Reactor Safety" Letter from Chairman, ACRS, to Chairman, NRC, dated May 16, 1979, I

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i Task II.E.2

(

May 1980 i

Subject:

" Interim Report No. 2 on Three Mile Island Nuclear Station i

Unit 2" Letter from R.' Fraley, ACRS, to Commissioners, NRC, dated April 18, 1979,

Subject:

"Recom.i.endations of the NRC ACRS Regarding the March 28, 1979 Accident at the Three Mile Island Nuclear Sta', ion Unit 2".

j Letter from Chairman, ACRS, to Chairman, NRC, dated April 7,1979,

Subject:

i

" Interim Report on Reactor Accident at the Three Mile Island Nuclear Station Unit 2" I

I s

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T II.E.2-6

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ML20034H273

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ML20034H273

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Dear Mr. Lochstet:

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