ML20078H203

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Reply to Util & NRC Responses to Limerick Ecology Action Severe Accident Risk Assessment Contentions.Certificate of Svc Encl
ML20078H203
Person / Time
Site: Limerick  Constellation icon.png
Issue date: 10/10/1983
From: Elliott C
LIMERICK ECOLOGY ACTION, INC.
To:
Atomic Safety and Licensing Board Panel
References
NUDOCS 8310140071
Download: ML20078H203 (69)


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'83 ECT 12 All:i1 UNITED STATES OF AMERICA 3

NUCLEAR REGULATORY CO.TiISSION Before the Atomic Safety and Licensing Board In the Matter of

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Philadelphia Electric Company

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Docket Nos. 50-352

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50-353 (Limerick Generating Station,

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Units 1 and 2)

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LEA'S REPLY TO APPLICANT AND STAFF RESPONSE TO SEVERE ACCIDENT RISK ASSESSMENT CONTENTIONS Preliminary Statement By order of SepteEber 21, 1983

(" Memorandum and Order Regarding Parties' Motion for Clarification and Schedule for Prehearing Conference"), the Board granted LEA the op-portunity to reply in writing to the ansvers of the Appli-cant and NRC Staff on the admissibility of LEA's proposed contentions regarding the environmental assessment of the impact of severe accidents.

This document constitutes LEA's reply.

LEA will first address the general concerns regarding admissibility of the contentions.

Integrated with this discussion is LEA's reply regarding Contention SARA-6.

Discussion of other contenticas follows, seriatim.

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SUMMARY

OF-ARGUMENT

.The National E nvironmental Policy ' Act (NEPA)l and 2

Commission regulations require a detailed' consideration of alternatives to a proposed licensing action.

The

. Staff is well aware of ongoing research into' measures to mitigate the impact of severe accidents, such as filter venting of containments, and other more reliable contain-ment heat removal' subsystems..Research on filter venting

=and other improved containment heat removal systems has I

been considered by the Commission in a proposed' policy I

statement to be sufficiently advanced to require the consideration of cost-effectiveness of'such measures in

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i Ifuture' construction permit applica'tions.3 To date, the environmental review of the Limerick g

licensing proposal has not considered any alternative to prevent or-mitigate the impact of severe. accidents.

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' LEA's contention SARA-6 contends that'NEPA and Commission

_ regulations implementing NEPA require some consideration in the licensing process of at least-those alternatives h

of1which'the staff is aware through Commission sponsored research and its review of the facility.

1 s

42 U.S.C.

s4321 et. seq.

2 10 C.F.R. 551.23 (c) 48 Fed. Reg. 16020

3

. t LEA's other SARA contentions specify certain in-adequacies of the environmental risk analysis performed

,A for NEPA purposes.

General Concerns Regarding Admissibility The Applicant-presents three arguments against ad-missibility of LEA's SARA contentions:

(1)

NEPA imposes no additional safety requirements in this proceeding.

(2)

No mitigative alternative to PECO's proposal for licensing is " feasible" or required.

(3)

The individual contentions are "with-out basis and specificity". 1 Applicant's summary cf its argument, "The Nuclear Regulatory Commission Has No Duty To Augment Its Safety Requirements.Under the Auspices of NEPA", demonstrates that the Applicant misses LEA's point, and ignores rel-evant Commission regulations.

While the obligations of an agency under NEPA are

" essentially procedural", (Vermont Yankee Nuclear Power 1

Applicant's responses to the individual contentions constitute nothing more than attempts to attack the merits of the, contentions.

As this Board is well aware, s

such attempts'.St this stage of the proceeding are im-proper (Houston Lighting and Power Co.,

(Allens Creek),

50-446, 11 NRC 542 (1980)), notwithstanding Applicant's concealing each argument on the merits in the veil of an attack on the " basis" for the contention. ~

A.

i Corp. v. Natural Resources Defense Council, Inc. 435 US 519, 558 (1978)), the purpose of those " essentially pro-cedural" obligations is to compel the agency to take a "hard look" a.

the environmental consequences of its actions (Cf. Kleppe v.

Sierra Club, 427 US 390, 410 n.

21 (1976), and to require it to undertake a " thorough study and a detailed description of alternatives" to the proposed action.

Monroe County conservation council, Inc. v. Volpe, 472 F2d. 693, 697-98 (2d Cir., 1972).

This requirement for a thorough study and a detailed des-cription of alternatives has been described as "the linch-pin of the entire impact statement." Id.2 NEPA requires more than merely the full disclosure of environmental consequences and project alternatives.

1 The National Environmental Policy Act requires an agency to include in every report on proposals for major federal actions significantly affecting the quality of the human environment a detailed state-ment on " alternatives to the proposed action". 42 U.S.C. 54 332 (2) (c) (iii).

See, 40 CFR 51502.14 (1980), considering the com-parison of alternatives to be at the " heart of an environmental impact statement".

10 CFR 551.23(d) requires that the Commission be guided by these CEO guidelines.

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s An-agency is not free to disclose adverse consequences of'a project, and alternatives to reduce those conse-quences and then completely ignore them.

NEPA requires their " full' consideration in agency decision-making".

-Natural Resources Defense Council, Inc. v. Grant, 355 F.

Supp. 280 (1973) [ Emphasis added].

While an agency need not elevate. environmental concerns over other appropriate considerations (Strycker's Bay Neighborhood Council v.

Karlen, 444 US 223 (1980)), an agency is not thereby' relieved of its obligation to consider alterna-tives to reduce. environmental impact.

The detail of the consideration must be' sufficient to show that the agency made a good faith effort to consider the values NEPA seeks to prot'ect by explaining fully the agency's course of inquiry, analysis and reasoning.

Philadelphia Council of Neighborhood Organizations v. Coleman, 437 F.

Supp. 1341 (E.D. Pa. 1977) aff'd without opinion, 578 F.

2d 1375 (3d Cir., 1977).

The concept of " alternatives" for NEPA purposes is indeed " bounded by some notion of feasibility", Vermont 1

Indeed, in Strycker's Bay, the trial court had found the agency's analysis of alternatives as " thorough and exhaustive", and the Court of Appeals had 6cnceded that the agency had given consideration to the alternatives.

At 226-227.

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Yankee Nuclear Power Corp. v. Natural' Resources Defense r

Council, Inc., 435 US 519 (1978).

However, "the' concept of ' alternatives'-i.s an evolving one, requiring the agency to explore more or fewer alternatives as they become better known'and understood." Id,. At 552-553

[ Emphasis added]

Tho obligation to consider alternatives is not im-

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posed solely by NEPA, butLis specifically required by the_ Commission's regulations:

The draft. environmental impact state-ment will include a preliminary cost-benefit analysis which considers and balances the environmental and other

effects of the facility and the alter-natives available for reducing or avoiding adverse environmental-and i

other effects...

i While satisfaction of-Commission stand-ards and criteria pertaining tc radio-logical effects will be necessary to meet the licensing requirements of the Atomic Energy Act, the cost / benefit analysis will, for the purposes of NEPA, consider the radiological effects of the facility and alternatives.

10 CFR 551.23(c)-

(Emphasis added]

Significantly, neither Applicant or Staff responses to SARA-6 even mention these Commission regulations im-plementing NEPA.

LEA believes that these straight-forward regulations and NEPA's clear language mandate some con-

sideration of mitigative alternatives.,

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.Yet rather.than conceding the point, both Staff

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and Applicant oppose the contention.

Applicant's arguments consist largely of_ counsel's assertions of

'the'"infeasibility" of alternatives which have not yet

'even been_ examined for Limerick.

Staff's filing pro-fesses ignorance of a matter which is the subject of an enormous research program sponsored by the Commission.

See, Proposed Commission Policy Statement on Severe Accidents and Related Views on Nuclear Reactor Regula-tion, 48 Fed. Reg.-16014, 16017-8.

The. thrust-of SARA-6 is simply that NEPA requires that the environmental review for Limerick reflect a consideration of the Commission's intensive investiga-tion'of mitigative alternatives and an application of that investigation to the_ Limerick facility.

This is a contention which addresses the threshold legal question of whether some consideration of mitigative measures is r'

required.

Both Applicant and' Staff in opposing the con-tention appear to contend that NEPA-does not require the i

crossing.of that threshold, and that the Commission is free to completely ignore even preliminary insights and findings produced by research into mitigative measures.

PECO specifically argues that (1) no mitigative alternatives are required under Commission regulations, and -(2) no mitigative alternatives are " feasible" for purposes of NEPA consideration.

These arguments improperly prejudge the outcome of any analysis of the alternativcs for Limerick, and ignore relevant Commission regulations, particularly 10 CFR 550.109, 10 CFR Part 50, Appendix A, and 10 CFR 551.23(c).

Applicant's suggestions that the Cm.rission lacks the regulatory authority to examine mitigative measures and direct their implementation if warranted by public risk, completely ignore 10 CFR 550.109, and the discussion in Appendix A concerning " additional criteria".1 The "backfitting" regulations of 550.109 specifically authorize the Commission to require design changes "if it finds that such action will provide substantial, additional protection which is required for the public health and safety or the common defense and security".

The Commission is further authorized to require a construction permit holder to " submit such information concerning the addition or pro-posed addition...or the modification or proposed modifica-tion, of structures, systems, components of a facility as it deems appropriate."

10 CFR 550.109.

1 SARA-6 is a NEPA contention, and does not assert that implementation of mitigative measures is necessarily required.

The decision to implement mitigation meas-ures which have been considered as NEPA requires is left to the Commission's descretion.

However, LEA's point is that Commission regulations amply authorize such a decision.

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Further, the Commission's regulations specifically contemplate -that "there will be some water cooled nuclear

-power plants for which the General Design Criteria are not sufficient and for which additional criteria must be identified and satisfied in the interest of public safety.

In particular, it is expected-that additional or different criteria will be needed to take into account unusual sites and environmental conditions..."

Part 50,-App. A.

While LEA has consistently taken the position that Limerick is a special case, it has also consistently argued

-that the Commission's. regulations themselves contemplate the identification of and appropriate treatment of such special cases.

Whether in fact, mitigative alternatives to PECO's licensing proposal are recuired for implementation at Limerick is a conclusion which must be reached in the course of examining.those alternatives, to determine the level of risk posed, and the opportunities for risk re-duction offered by alternatives.

Applicant's assertions of " remoteness" and "infeasibility" merely prejudge the outcome of the analysis.

The Indian Point licensing board has recognized that such matters cannot be " prejudged".

i With respect to contentions advocating the necessity for filtered vented containments or separate containment struc-i I

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tures for relief _of excess pressure at the Indian Point

" facilities, the Licensing Board rejected the licensees' arguments that prior to admitting the contentions, the Board is first required to determine that a significant risk exists without those measures:

l' "We do not believe that the Commission intended that prior to admitting a con-l tention advocating a safety measure, we i

should find that a significant risk sure-

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ly exists without such safety measure.

j We believe such a finding should reflect the outcome of this litigation rather than its starting point..." In Re Consolidated Edison Co. of New York (Indian Point)

Dockets 50-247, 286-SP, 16 NRC 1629 (1982) at 1634.

[ Emphasis added]

l Similarly, PECO's arguments that no mitigative alternatives need be considered because NEPA does not require consideration of " remote and speculative" al-ternatives (PECO pleading, p. 9-11) prejudge the out-come of an examination of alternatives.

A fortiori, one must look at what alternatives exist before one determines them to be " unfeasible".

The prematur'ity at this stage of the proceeding of PECO's arguments on l

the merits is well illustrated by Houston Lighting and l

Power Co.,

(Allens Creek Nuclear Generating Station)

I Docket 50-446, ALAB-590, 11 NRC 542 (1980) (reversing i

the Board, and ordering it to admit a contention that a 256 square mile biomass farm was an environmentally f

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preferable and viable alternative to the

_ lens Creek facility).

Obviously, the mitigative measures contemplated by LEA's contention are nowhere so " remote and speculative" as.the-256 square mile " biomass farm" that the Appeal Board held to sufficiently delineate an " alternative" for purposes of a NEPA litigation contention.

Among the mitigative measures contemplated by the contention are a filter venting of the containment.

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alternative, which Applicant characterizes as so remote and speculative that it need not even be examined for NEPA purposes, is being required in a simplified form by France's Commissariat a l'Energie Atomique, for im-

plementation within 2 years at all French nuclear facil-ities.

See, "Inside N.R.C.",

Vol.

5, No. 18, (September 5, 1983).

In the United States, programmatic analysis of vent-filtered containments is well under way.

See, e.g.,

NUREG/

CR-1029, SAND 79-1088, " Program Plan for the Investigation of Vent-Filtered Containment Conceptual Designs for Light Water Reactors" (Sandia Labratories, October 1979).

Another alternative which the contention suggests and - -...

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- which is being investigated are various options for core retention.

See, e.g., NUREG/CR-2155, SAND 81-0416, "A

Review of the Applicability of Core Retention Concepts to Light Water Reactor Containments", (Sandia Laboratoreis, September, 1981).

As a result of its ongoing research program into mitigation strategies, the Commission has already deter-

. mined that some of the mitigation alternatives which PECO argues are too remote and speculative for NEPA consideration are sufficiently feasible to warrant detailed cost-ef fec-tiveness' analysis:

"In future CP applications for both pressurized water reactors [PWRs ]

and boiling water reactors (BWRs ]

filtered-vented containment systems or a variation of such' systems, should be provided if these yield a cost-effective reduction in risk."

Proposed Policy Statement on Severe Accidents and Related Views on Nuclear Reactor Regulation, 48 Fed. Reg. 16014 (April 13, 1983) at 16019.

So, too with more reliable containment heat removal:

"The Staff is studying the need for more reliable subsystems for contain-ment heat removal...as possible al-ternatives to filtered venting for prevention of gradual overpressuri-zation failure of the containment building.. The cost-effectiveness of this alternative should be con-sidered in the design of plants for new CP application."

Id.,

at 16020. 2

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'These references establish an adequate factual basis

.for LEA's contention that adequate. cost-effectiveness re-

'viewLis possible for NEPA' consideration of these mitigative alternatives.for the Limerick ~ facility.

PECO_ complains at length about uncertainties asso--

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ciated with probabilistic risk assessment techniques, and asks _this: Board to conclude (prematurely, at this early stage'of determining contention' admissibility, without any record upon which such a conclusion could be reached) that "it is presently inappropriate, and as a practical matter, impossible to utilize these [PRA } techniques to impose-additional: safety-requirements for this facility under the auspices of NEPA"'.

.PECO' Pleading, at p.

6.

This-prejudgment has been rejected in the Commission's proposed policy statement

"In sum,Lconsidering the experience with risk assessments thus far made, we conclude.that the cost-effective-ness of risk reduction measures-can be studies through PRA.

Although

~there are limitations.due to the

'many uncertainties associated with the use.of PRA,'the Commission-con-siders it'to be'a valuable adjunct to-the' established regulatory pro-cess and the NRC's nuclear safety regulations in 10 CFR, Chapter I".

Proposed Policy Statement, at 16016..-

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i It may, or may not be, upon a consideration of alternatives, their cost,'and their potential for risk reduction, that. construction has proceeded so far that no mitigative measures will be cost-effective for Limerick.-

That, however, is a conclusion which can only be reached by the Staff after it has considered the rel-evant factors.

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In contrast to the Applicant's prescience of the results of an alternatives analysis, the Staff scarcely.

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'l Applicant also argues that "only alternatives which are feasible in the time frame of facility licensing need be considered", but offers absolutely no author-ity for such a proposition.

To the contrary, NEPA

. requires a consideration of alternatives which are available in a time frame which is " meaningfully compatible-with the time-frame of the needs to which the' underlying proposal is addressed".

See, Vermont Yankee, supra, at 551, citing Natural Re-Sources Defense Council v. Morton, 458 F2d 827, 837-838 (1972).

In this case, the operating li-cense.for Limerick is for a duration of 40 years

.and.the facility is expected to meet a-need for power over'that time period.

LEA submits that NEPA requires some consideration of alternatives which would be available within that 40 year period.

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i seems to know what LEA is talking about at all, despite

- the fact that mitigative measures for plants including Limerick are the focus of intensive scrutiny of Commission i

technical research and regulatory debate.

The Staff response to SARA -6 complains that "no risks are identified at all" by' LEA, and "no alternatives and mitigative measures are proposed" and that LEA is "not raising a contention but calling for an unspecified Staff study".

(NRC Staff, Pleading, p.8)

Let us recall that we are discussing contentions

. filed in response to a document that for two volumes discloses the risk of severe accidents at Limerick and which has been filed in this proceeding.

-The " risk" to which the contention refers is ob-viously the risk of severe accidents as disclosed by SARA, and by the BNL report to which the LEA's filing specifically referred.

'As far as " proposing" alternatives and mitigative

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measures is. concerned, any such proposal is premature in view of the Staff's refusal to concede the threshold legal question to which the contention is directed.

In any event, the Staff-is well aware of the existence of mitigative l

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alternatives for Limerick, having sponsored a preliminary analysis of such measures.

See, e.g., NUREG/CR-2666, Chapter 7, "Further Considerations of Mitigative Features for Specific Plants:

Limerick", PWR Severe Accident Delineation and Assessment (attached).

It appears, in fact, that while the Staff is contesting the NEPA necessity for an alternatives analysis, it is con-ducting-just such an' analysis for other purposes.

' Subsequent to the original filing of LEA's SARA s

contentions, pursuant to the Freedom of Information Act,.

LEA sought l disclosure by the Commission of non-public records relating to mitigation features for Limerick.

The Commission finally responded to this request on October 3, 1983,'providing to LEA's counsel information which hereto-i

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fore was unavailable to LEA.

That information clearly establishes that tho. Staff has received significant infor-mation concerning mitigation strategies for Limerick:

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Summary to date: ~.For Mark II contain-ment as exemplified by the Limerick Plant, mitigation requirements (functions) have baen identified, in-cluding containment heat removal, core residue capture and retention without concrete attack, and (if ATWS events are to be mitigated) some' kind of vent-ing system.

Candidate components to fulfill these requirements have been selected for preliminary conceptual design and cost estimation.

Separate. _ _

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. V cost figures will be generated for
1) Plants before construction begins, 1!) Plants built but not yet in op-eration, and 3) Operational plants.

d.

Plans for next period:

Complete pre-liminary designs and assessments for Limerick,'and begin final design of selected version.

" Monthly Project Status Report, September 15, 1983, Contract NRC-03-83-092",ly( R&D Associates, p.

4.

A copy of some of the information provided to LEA pursuant to its request is attached hereto, and provides '

additional factual bases for SARA-6.

.This information concerning feasibility of mitigation alternatives which is available to the Staff is surely sufficient to " require reasonable minds to inquire further" (Vermont Yankee, supra, at 554), into this alternative for NEPA consideration.

In fact, the " unspecified Staff study"

-(NRC Staff, Pleading, p.8) which LEA has contended is required for NEPA purposes is apparently being performed for other Commission purposes.

The Staff knows exactly what analysis is required.

The question is whether what the Staff already knows must be' considered in this licensing proceeding for purposes of complying with the NEPA requirement of a detailed consider-ation of alternatives.

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4 SARA-1 and 2 Both the Applicant and Staff object to these contentions on

'the grounds that they were previously submitted as PRA contentions i

and' denied by this Board.

LEA resubmitted these contentions as SARA contentions, in spite'of.the Board's previous ruling, because LEA read that ruling to hinge primarily on the Board's decision not to litigate PRA methodology for safety purposes.

LEA did not

request reconsideration of the Board's ruling on these two contentions due to that fact.

LEA has'not attempted to'" repackage" the contentions by making them more specific, since it would have been very difficult to do so without reproducing large portions of the Brookhaven report (NUREG/CR-3028) ; the Board is thus faced with a straightforward decision as to whether or not the contentions can be resubmitted for a different purpose

-than-they were originally submitted.

If they were denied the first time due largely to the Board's decision regarding litigation of PRA methodology, there is no problem with their reconsideration in a SARA (NEPA) context.

If the Board is not convinced, based on LEA's statements regarding specificity, that the contentions are sufficiently specific to admit, they will again be denied..

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SARA-3 Applicant objects to this contention alleging that LEA:

erroneously' assumes that the entire population between 10 and 25 miles frem the facility will be relocated.

There is no basis given by LEA for asserting that such total evacuation would be necessary or was assumed in the model.

In~ fact, the model simply

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assumed that the portion of.the popu-lace which could be affected by depo-

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sition from the plume would be re-located-once the plume'had passed.

PECO Pleading, p. 21.

This objection impermissibly addresses the' merits, not the basis of the contention, and is based upon counsel's assertions nowhere supported by any matter appearing in SARA.

In any event, the objection. misses the point of the con-tention.

Even assuming the truth of Applicant's counsel's i-

-unsupported assertions that_"only one, or at most a few of the 16 sectors at these distance would require relocation",

the problem addressed by the contention still remains.

SARA shows that-the year 2000 population of sector SE is 680,330.

(SARA,p. 10-33)

The year 2000 population in the' sector.ESE.is 505,011. (SARA, p. 10-33)

If "only one Anc at most a'few" of the sectors require rapid relocation,

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-.---_a the total population to be " rapidly relocated" still'may easily exceed one million people.

NRC. Staff has under-stood the point:

" LEA's point is that the area outside of the ten mile zone contains a high population density and asserts that for this reason the assumotion in SARA, that people in-the zone from ten to twenty. miles from.

Limerick can be rapidly relocated, has'an inadequate factual foundation".

(NRC Staff Pleading, p. 5)

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SARA-4 The-Applicant's objection _to this contention once again consists of a defense on the merits which properly belongs in hearing testimony, not in.an objection to contention admissibility.

Contrary to Applicant's assertions that LEA " points to no better-data which was overlooked in determining the

- best way to model evacuation" (PECO Pleading, p.22) the contention specifically points to-the site-specific high-

- way movement time of the~ Emergency Plan Appendix H evacuation-time estimates.1 Applicant complains about cEhe fact that Appendix H "does not take into account-the

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- prompt' notification system", and the impossibility of separating'" preliminary evacuation notification time from preparation time'which makes such data difficult to use in SARA" (PECO Pleading, p.23).. But LEA never proposed the use of' Appendix H's notification and preparation time 1

'While' pointing to this' data as "botter", LEA does not necessarily propose its use until the errors delineated in LEA Contention VIII-6 are corrected. _

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estimates.

The contention proposes using only the highway movement time component (a component iddepend'nt of e

notification-time) from'which~an evacuation speed can be calculated.

Delay time (including preparation and noti-fication time) is calculated separately from evacuation speed in the CRAC 2 evacuation model which SARA incorporates, and-the proper evacuation ~ speed can easily be entered as input in the CRAC 2 evacuation.model.

With respect to the 3-hob: notification time for seismically' induced accidents, Applicant asserts that the low evacuation speed assumed "would compensate for any shift in notification time".

(PECO pleading, p. 23).

This assertion once again belongs in hearing testimony.

Ecwever, LEA notes that it has earlier noted in this proceeding that health effects calculations are extremely sensitive to evacuation delay assumptions (See e.g.,

LEA Filing of April 12, 1983 (Spec-ification of Conditionally: Admitted' Contentions, I-16b)),

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and the low evacuation. speed would not necessarily compen-sate. L

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..o Appiicant objects to. SARA-5, alleging that "many of the

. consequences which are alleged to be omitted are, in fact, directly.or' indirectly included in SARA.

LEA's complaint

-seems to be that instead of or in addition to expressing

', consequencss' in' terms of dollars, SARA should state effects in terms of other units of measure, e.g.,

acres'of land or job losses".1 (PECO Pleading, p.24)

To tNe contrary,. LEA contends that NEPA requires a dis-closure f what the consequences will be - not their obsfucation of the consequences by disclosing only their economic impact.

Acceptance of Applicant's reasoning would permit an EIS to express all environmental impacts solely in economic' terms.

Such descriptions would conceal,

-not disclose, environmental-impacts.

The impacts should.be stated in terms that describe the

-impact - if 10 square miles of land will be permanently

~ interdicted due to contamination, this impact should be so stated.

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While~ Applicant provides.a list of types of impact that are " included in CRAC 2 runs" (PECO Pleading, p.25) they are not set'forth in SARA.

When LEA's counsel requested certain results printouts from the CRAC 2 runs for

SARA, Applicant's counsel informed him that the only runs done were those set forth in SARA.

If Applicant is willing to include in SARA complete CRAC 2 final'results printouts, LEA's contention as to matters (3)

(6) would be satisfied.

With respect to health effects, while SARA does consider total thyroid nodules-(SARA, p.12-17), the contention seeks disclosure of the total of all non-fatal cancers (not merely thyroid nodules) and other health effects known to be associated with radiation exposure such as i

sterility and genetic effects.

Applicant does not even address these omissions.

(

With respect'to items in (14), Applicant-admits that SARA does not consider them " individually", and that the CRAC 2 runs which "implicily " consider them rely upon national i

averages.

Once again, only-the economic impact is consider-t ed; _further, " national averages" do not reflect site-specific nature of the mineral and water resources and scenic and i

aesthetic resources referred to..

With respect to the compensation cost of health effects, LEA sees no reason why these costs should not be treated similarly to, and included within, the CCDF curves of "ex-plant costs" set forth in SARA.

SARA's different t

treatment of such costs obscures their significance, contravening NEPA's " full disclosure" purposes.

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s SARA-6 LEA's reply to objections to this contention are integrated into its general discussion regarding admissibility of SARA contentions.

LEA corrects, for the record, a typographical error in SARA-6.

At page 20 of its SARA contentions, the last line should read:

Hampshire, (Seabrook Units 1 and 2), 6 NRC 33 at 83.

It previously read:

Hampshire, (Seabrook Units 1 and 2), 5 NRC 33 at 83.

SARA-7 Both the Applicant and Staff rely on the Commission's Safety Coal Policy Statement of March 14, 1983, for the proposition that sabotage cannot be considered as part of the risk assessment for NEPA purposes.

I LEA believes that a distinction should be made between the analysis needed for setting an absolute safety goal and the analysis that could be done for NEPA purposes.

Sholly lays out in his discussion the information available for making such a risk assessment for NEPA.

LEA believes that it is the duty of both the Staff and Applicant to make the best assessment presently possible of the risk of sabotage, recognizing that such an analysis will carry with it uncertainties sufficient j

to justify excluding it from the safety goals at this time.

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Applicant also objects to this contention on the basis that Mr. Sholly has not been shown to be an expert in the field of probabilistic risk assessment.

LEA notes for the record that in the recent special investigation at Indian Point, Mr. Sholly was recognized by Mr. Gleason, the Chair for that investigation, to be qualified to testify with respect to a wide variety of matters concerning nuclear safety.

He was in fact qualified in that proceeding to testify as an expert on consequence analysis.

Both Staff and Applicant object to SARA-7 (b).. LEA notes that this contention was submitted as a PRA contention previously o

(I-23c), and was held in abeyance by the Board for possible refiling as a SARA contention to whatever degree necessary.

It was reworded as resubmitted in order to comply with the Board's order that: parties with similar contentions attempt to agree on wording prior to submission -- the City of Philadelphia brought into discussions with LEA a contention regarding errors of commission.

That requirement necessitated keeping the wording of joint contentions simple.

ly submitted, Respectf

70. Ltlant LLluu Charles W.

Elliott fhN Judith A.

Dorsey Counsel for Limerick Ecology Action l !

x.

~ - ~. _

CERTIFICATE OF SERVICE Ej.;f IherebycertifythattheforegoingLEA'sReplytg3qg[jgaggq)

A and Staff Response to Severe Accident Risk Assessment Contentions LFF;u was served on October 10, 1983,byfirst-classmail,p5Et'Egym.h,?l d

prepaid, upon the following*:

Lawrence Brenner, Chairman Atomic Safety and Licensing Board Administrative Judge Panel U.S. Nuclear Regulatory U.S. Nuclear Regulatory Commission Commission Washington, DC 20555 Washington, DC 20555 Atomic Safety and Licensing

  • Dr. Richard F.

Cole Appeal Panel Administrative Judge U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Washington, DC 20555 Commission Washington, DC 20555 Docketing and Service Section Office of the Secretary

  • Dr. Peter A. Morris U.S. Nuclear Regulatory Commission Administrative Judge Washington, DC 20555 U.S. Nuclear Regulatory Commission Zori G.

Ferkin, Esc.

l Washington, DC 20555 Commonwealth of PA Department of Environmental Resource Ann P. Hodgdon, Esq.

505 Executive House Office of the Executive P.O.

Box 2357 t

j Legal Director Harrisburg, PA 17120 U.S. Nuclear Regulatory Commission David Wersan, Esq.

Washington, DC 20555 Assistant Consumer Advocate Office of the Consumer Advocate

  • Jessica H. Laverty, Esc.

1425 Strawberry Square Conner and Wetterhahn Harrisburg, PA 17120 1747 Pennsylvania Ave., NW Washington, DC 20006 Director PA Emergency Management Agency Phila. Electric Company Basement, Transportation and ATTN: Edward G. Bauer, Jr.

Safety Building VP and General Counsel Harrisburg, PA 17120 2301 Market St.

Phila., PA 19101

  • Sent by Express Mail

..s.----

- ~ -...

Thomas Gerusky, Director Spence W. Perry, Esq.

Bureau of Radiation Protection Associate General Counsel

(

Department _of Environmental FEMA Resources Rcom 840 Fulton Bank Building, 5th fl.

500 C St.,

SW Third and Locust Sts.

Washington, DC 20472 2Harrisburg, PA 17120 6

Martha W. Bush, Esq.

Angus Love, Esq.

Deputy City Solicitor 101 East Main St.

City of Philadelphia Norristown, PA 19401 Municipal Services Building 15th and JFK Blvd.

Jay M. Gutierrez, Esq.

Phila., PA'19107 U.S. Nuclear Regulatory Commission, Region 1 Robert Anthony 631 Park Ave.

103 Vernon Lane, Box 186 King of Prussia, PA 19406-Moylan, RA 19065 Marvin Lewis 6504 Bradford Terrace Phila., PA 19149 Jacqueline I.

Ruttenberg Keystone Alliance 3700 Chestnut St.

Phila., PA 19104 Frank Romano

.61 Forest Ave.-

l Ambler, PA 19002 Joseph H. White III 8 North Warner Ave.

3ryn Mawr, PA 19010 Robert Sugerman, Esq.

Sugarman and Denworth Suite 510, North American Building 121 S.

Broad St.

i Phila.,-PA 19107-l 1

3

.s A)))

G N -

U/

JUQITH A.

DORSEY

/)

1/

U E.-

N.,

DOCKETED US E CHAPTER 7.

FURTHER CONSIDERATION OF MITIGATION FEATURES F@ SfECI,F,ICril M2 PLANTS: LIMERICK As part of the review and analysis of the Probabilistic Risk Assessment (PRA) for Limerick [1], the question of mitigation features was considered.

This

{

cc* sideration is divided into two sections.

In Section 7.1, mitigation is I

(

considered within the context of the PRA as submitted.

Because of possible changes due to updating, inconsistencies and further analysis of containment behavior, Section 7.2 considers mitigation in terms of alternative scenarios.

f In both cases, the discussion is qualitative. Section 7.3 has concluding

[

remarks.

7.1 THE LIMERICK PRA The dominant accident sequences identified in the Limerick PRA were grouped 1

into four classes, as shown in Table (7.1).

Note that Classes II and IV involve containment failure prior to core melt, and Classes III and IV involve transients with loss of scram function (ATWS). The containment event trees identified eleven potential failure modes, which were eventually combined into.

seven failure modes, as shown in Table (7.2). The frequencies of each of these seven failure modes for the four accident classes are shown in Table (7.3).

Although Table (7.3) implies that 28 release categories pou'Id have been used in the consequence analysis, only five were used.

These are shown in Table 9

(7.4), which also indicates how the containment failure modes were. combined into the five release categories.

For the consideration of mitigation, it is important to note the following.

Release category OXRE combines all the in-vessel and ex-vessel steam explo-i 7-1

TABLE 7.1.

Accident Sequence Classes Generic Accident Physical Basis System Level Contributing Sequence Designator for Classification Event Sequence t

Class I (C1)

Relatively fast core melt; Transients involving loss of inventory containment intact at core makeup, small LOCA events involving loss melt and at low pressure of inventory makeup Clacs II (C2)

Relatively slow core melt Transients or LOCAs involving loss of due to lower decay heat heat removal, inadvertent SRV ' opening power; containment failed accidents with inadequate heat removal prior to core melt capabili,ty As Class III (C3)

Relatively fast core melt; Transients involving loss of scram function containment intact at and inability to provide coolant makeup, core nelt, but at high large LOCAs with insufficient coolant internal pressure makeup transient with loss of heat removal and long term loss of inventory makeup Class IV (C4)

Relatively fast core melt; Transients involving loss of scram function containment fails prior to and loss of containment heat removal or all core melt due to over-reactivity control, but which have coolant pressure makeup capability I

TABLE 7.2.

Containment Failure Modes CONTAINMENT FAILURE MODES Designator Description Steam explosion in vessel a

S Steam explosion in containment p'

H exp1 sion induced containment 2

failure p

H deflagration sufficient to 2

cause containment overpressure failure 6

Overpressure gmall leaks (AR = 0.05 ftc) 2 y

Overpressure failure (AR = 2.0 ft )

Release through drywell 2

d y'

Overpressure failure (Ag = 2.0 ft )

Release throLgh wetwell break 2

y

Overpressure failure (Ag = 2.0 ft )

Wetwell pool drained Overpressure $)largeleak c

(AR = 0.2 ft Overpressure,large}eak,SGTS cc failure (A = 0.2 ft )

i R

de Overpressure, small leak, SGTS 2

failure (AR = 0.05 ft )

N e

w W

I U, #.

'(

's i 7 -3

t is i

TABLE 7.3

SUMMARY

-- GENERIC ACCIDENT SE00ENCE/ RELEASE PATil COMBINATIONS

  • CLASS TOTAL PROBABILITY CONTAINMENT BY CONTAINMENT FAILURE N0DE CLASS I CLASS II CLASS III-CLASS IV FAILURE MODE

-8 a

1.2x10 9.6x10-10 1.1x10-9 1.3x10-10 1.5x10-8 S,p' 2.5x10-8 1.9x10-2.2x10-9 2.5x10-10 2.9x10-8 3.2x10-6 2.5x10-7 2.8x10 6.4x10-8 3.8x10

-7

-6 y,p y'

2.8x10-6 2.1x10-7 2.4x10-7 5.6x10-8 3.3x10-6

}

y

3.1x10-7 2.4x10-8 2.7x10-8 6.3x10 3.7x10-7

-9 cc,6c 9.7x10-7 7.5x10 8.5x10 2.5x10-II 1.1x10-6

-0

-8

-6

-7

-II c,6 5.2x10 4.0x10-7 4.6x10 2.5x10 6.1x10-6 TOTAL PROBABILITY 1.2x10-5 9.6x10-7 1.1x10-6 1.3x10-7 1.5x10-5 BY CLASS

  • Reproduced from Table 3.5.14 of the Limerick PRA

TABLE 7.4.

Release Categories Used in the Limerick PRA Release Containment Categories Failure Modes t,

C4y y' - Class IV i

C4y' yu - Class IV C4y" y" - Class IV f

a - Class I, Class II, 1

Class III, Class IV, 0XRE f Su' - Class I, Class II, Class III, Class IV.

'y

- Class I, Class II, t

'i l

Class III l

OPREL j

l f y' - Class I, Class II, i

Class III f y" - Class I, Class II, to Any Release Class III, Category

i. (c,6c - All Classes t

E,6

- All Classes 4

7-5

sions, and hydrogen detonation failure modes.

Release category OPREL combines drywell and wekwell overpressurization failures for Classes I, II and III.

Over 50% of the total probability is not allocated to any release category.

Included in this percentage are overpressurization failures of the wetwell (assuming that the suppression pool drains) for Classes I, II and III, and all leakage failures.

However, based on th.e five release categories, Philadelphia Electric Company (PECO) [2] was able to produce the risk as a function of release category These are shown in Tables (7.5) and (7.6). The calculations carried out by PECO show that, with reference to latent fatalities as a consequence measure, the total risk is distributed among the five release categories in the following way:

Release Percentage of Total Category (latent risk)

C4y 1.49 C4y' 2.35 C4 "

0.34 7

0XRE 4.02 0PREL 91.79 By the use of Table (7.4), it is possible to redistribute the risk directly among failure modes and classes. The result is shown in Table (7.7).

Tnis table can be used as the basis for a discussion on risk mitigation. Q apparent that more than 90% of the risk is associated with overoressure M1 ore (with either wetwell or drywell break).

Therefore, strategies for mitication have to address this failure mode.

m

e TABLE 7.5.

ACUTE FATALITIES f

(TOTAL RISK)

NORMALIZED MEANS RELEASE PROBABILITY MEANS BY PERCENTAGE CATEGORY (from Table 7.3)

( from PECO).

CATEGORY OF TOTAL C4y-5.6 (-8)*

3.556 0.199 (-6) 8.5 C4y' 6.4 (-8) 11.02 0.705 (-6) 29.9 C4y" 6.3 (-9) 75.78 0.477 (-6) 20.3 OXRE 4.35 (-8) 22.37 0.973 (-6) 41.3 OPREL 6.98(-6)

TOTAL 2.35 (-6) 100

  • 5.6 (-8) = 5.6 x 10-8

-i 1

}

7-7

TABLE 7.6.

LATENT FATALITIES (TOTAL RISK)

NO W LIZED PROBABILITY RELEASE MEANS MEANS BY PERCERTAGE CATEGORY (from Table 7.3)

(from PECO)

CATEGORY OF TOTAL C4y 5.6 (-8)*

93.26 0.052 (-4) 1.49 C4y' 6.4 (-8).

127.8 0.082 (-4) 2.35 C4y" 6.3 (-9) 182.8 0.012(-4) 0.34 OXRE 4.35 (-8) 332.0 0.14 (-4) 4.02 OPREL 6.98 (-6) 46.38 3.196 (-4) 91.79 i

TOTAL 3.48 (-4) 100.0

  • 5.6 (-8) = 5.6 x 10-8 l

L t

I~

L sLe

'Before considering the possibility of backfitting a Filtered-Vented Contain-ment System (FVCS)', however, it is necessary to know the possible role that

~

containment sprays.can have in case of accident. _ Containment overpressure is due.to both steam and condensible gases.

If, as is usually the case, steam is-the dominant contributor to overpressure, then containment sprays would effectively: reduce containment pressure in~certain accident sequences.

Hcwever, because of the uncertainties on the implementation of containment sprays under postulated severely _ degraded core conditions, their effect was not evaluated in the Limerick PRA.

It would, therefore, be interesting to quantify the mitigation effect that the sprays can have in some sequences.

Consideration should also be given to the possibility of upgrading their performance in order to cope with the severe environmental conditions in case of accident.

'If the sprays do not.seem to reduce the risk significantly, then a FVCS can be considered as a mitigation feature.

Work done at Sandia for Mark I and Mark III containment by A.S. Benjamin and F.T. Harper [3], has 'shown that, despite the differences between the two con-tainments, basically the same filtered-venting strategies are thought to be

. effective in reducing the total risk significantly.

l

[

From the viewpoint of mitigation by a FVCS, Table (7.7) can be reorganized in l

l.j the following way:

if the r and p containment failure modes' contribution i= I are decoupled by the use of the containment event trees, the following results:

L I

{

i E

7-9 I

TABLE 7.7.

Total Risk Contributors Containment Failure Mode Class :

I II III IV c

1.11 0.09 0.10 0.01 1.31 S, v' 2.30 0.18 0.20 0.02 2.70 y, p 42.09 3.29 3.68 2.35 51.41 y'

36.83 2.76 3.16 1.49 44.24 0.34 0.34 y"

82.33 6.32 7.14 4.21 100.00 From Table 7.5, it is clear that release category OXRE is an important con-tributor to acute' fatalities, even though OPREL has the highest probability.

This occurs because the Limerick PRA consequence analysis determined that OPREL would result in zero early fatalities.

Since this analysis is under review by the NRC Staff and their consultants, the discussion here will focus on mitigation of the y and y' failure modes, using latent fatalities i

as the measure of risk.

If acute fatalities are excluded, it would appear l

that overpressurization of drywell or wetwell for Class I accidents is the largest contributor to risk.

7-10

Release Category Percentage of Total A - non-mitigable failure modes:

a, S, p', u 5.86 8 - mitig(able *, no ATWS: y, y classes I, II) 83 34 C - miticable

  • ATWS:. y, y ', y"

(' classes III, IV) 10 80 100.00 Therefore, a low-volume venting strategy can potentially eliminate contribu-tion B and the risk reduction factor is %6.

On the other hand, a combined low-volume /high-volume strategy, similar to the one proposed by Benjamin for a Mark I BWR, can potentially eliminate contributions B and C and the risk reduction factor is %17. This combined strategy may be necessary because ATWS events tend to give higher loadings over short periods of time.

Category A, above, assumes that a hydrogen burn will fail containment with a probability of one. Since this is conservative, the risk reduction factor may even be greater.

  • Note that for Class II and IV, the containment fails prior to core-melt.

Hence only a vent might be required.

7-11

~

7.2.

LIMERICK MITIGATION The Limerick PRA [1] expressed ideas about how vessel failure occurs and how subsequent dispersal of the core materials led one to the conclusion that N cs failure fairly early in time would lead to immediate containment 9

N failure.

Most of the dangerous decay product 3 were assumed to have been sparged by the suppression pool before vessel failure.

Currently it is thought that vessel failure will lead to core materials entering the suppression pool within the pedestal.

It is argued that the suppression pool will act as a heat sink of sufficient magnitude to condense the initial large amounts of steam.

The result is a slow overpressurization and containment failure, be it either the y or the y' mode.

It is not clear why the y" mode is considered improbable in the Limerick PRA.

The risk is not very different because of the early blowdown through the SRVs and resultant sparging by the suppression of the more dangerous constituents. What the fraction of the more dangerous decay products that are transported through the SRVs should be is addressed elsewhere.

(

The earlier sequence that leads to containment failure caused by diaphragm failure does not lend itself easily to mitigation.

The diaphragm must be l

protected.

To do this, we might consider modifying the region under the l

l reactor vessel, if existing drains do not supply a sufficier.t path, so that the core materials enter the suppression pool without failing containment.

This might be accomplished by replacing part of the concrete floor with an easily melted metal cover.

Ccmments about mitigation of an event such as this would be the same as for the current scenario which is discussed below.

l l

7-12

r The slow overpressurization that is believed to follow core material entering the suppression pool could be mitigated by heat removal from the containment.

This could be accamplished by a low volume flow vent-filtered system, a heat pipe or a cont:inment spray system. Heat pipes or a vent filter have been discussed elsewhere in this report for application to PWRs.

The BWR Mark II containment is similai to a PWR when cavity flood precedes vessel failure.

A spray system could mitigate slow overpressurization accidents, providing suppression pool cooling is possible.

To increase reliability, a system could easily be devised that could be driven from outside the containment, using equipment brought to the site from elsewbre (portable pumps, diesels, etc.). Some consideration should be oiven to genHng the snrav system as an alternative method of mitication.

It should be kept in mind that, if water is added to the containment, it must eventually be removed.

Some sort of a closed loop heat exchange process seems to be the most desirable approach.

PECO and their consultants [4] give arguments as to why slow pressurization will follow direct entry of the molten core material into the suppression pool.

The PECO view is that very rapid condensation will lead to very little pressure increase due to steam generation and no communication of the wet well with the dry well.

Questions have been raised regarding the efficiency of the condensation process.

Calculations, assuming zero condensation, lead to very early containment failure.

It is our opinion that the condensation rate will not be infinitely fast.

The high rate of steam generation--recall the PWR steam spike--will overwhelm heat pipes.

One is led to consider a high

_ volume vent-filter system opprays of very high capacity (ifyrated in a-timely manner).

This area needs a great deal more attention before conclu,sive A

can be madef 7-13

~

Whether or not the core debris bed will remain coolable within the pedestal needs to be det smined.

If there is any doubt, a rubble bed within the pedestal may be desirable. Suitable flow passages in the pedestal wall at the suppression pool floor level are needed.

At this time it is not clear that they exist.

Debris bed dryout is discussed in NUREG 0850, and the rubble bed and its design are discussed by Swanson [5].

Further consideration should be.given y" failure of the containment because it would make mitigation much more difficult.

The highly contaminated suppression pool water would flood surrounding areas and could create no great difficulties. Mitigation would require prevention of the y" mode.

As this mode of containment failure is potentiallv the most dangerous, it e

deserves serious attention.

7.3 Concluding Remarks The PRA and its ammendments, as submitted by PECO did not give realistic or best estimate scenarios.

Rather, some conservative as well " mntro-versial assumotions were made.

One can only consider mitigation in the con-text of realistic, best estimate accident and containment failure scenarios.

l It is concluded that the next step should be a better definition of these scenarios so that mitigation strategies can be developed if proved to be l

necessary.

REFERENCES

[-

1.

Philadelphia Electric Company, " LIMERICK Generating Station Probabilistic Risk Assessment,1981.

2.

Letter, Treavor Pratt to W.E. KastenLerg, August 1982.

3.

A.S. Benjamin, F.T. Harper and P. Cybulskis, " Risk Assessment of Filtered-Vented Containment Options for a BWR Mark I Containment",

Proceedings of the International ANS/ ENS Topical Meeting on Probabilistic Risk Assessment, September, 1981.

7-14

--t-

., i-e n w

y.-

v-w-i.-

e--

e s-<

ww-y-,

r,-----r'

4.

Summary of Meeting Between PECO and Consultants and NRC and Consultants, 3 Sept.1982, to be published.

5.

D.G. Swanson, " Core' Melt Materials Interaction Evaluation", Annual Progress Report for April 1980.to March 1981,_ ASAI Report No.81-001.

A L

l 7-15

4*.. ~,%

[

UNITED STATES J'8 N

'?

NUCLEAR REGULATORY COMMISSION 3'

E WASHINGTO N, D. C. 20000 00NETEP

,/

usNR0 October 3,1983

'83 CCI12 All M2 7.,

Charles W. Elliott, Esquire

- N. N

-Brose and-Poswistilo 1101 Building IN RESPONSE REFER Easton, PA 18042 TO F0IA-83-550 -

Dear Mr. Elliott:

This is.in response to your letter dated September 13, 1983 in which you requested, pursuant to the Freedom of Information Act, drafts of Volume 2 of NUREG-0850 and documents regarding the feasibility of measures or design features to mitigate the consequences of a core melt accident at Limerick or any other reactor of the GE Boiling Water Mark I, II, or III Design.

During a September 23, 1983 telephone conversation with Mr. Stephen Isaacs of my staff you, (1) eliminated all Mark III design material from the request, and (2) were informed that there is no draft or manuscript of NUREG-0850, Volume 2.

The following documents address mitigating features in Mark-II containments:

1.

NUREG/CR-3028, "A Review of the Limerick Generating Station Probabilistic Assessment."

2.

NUREG/CR-2666, "PWR Severe Accident Delineation and Assessment."

3.

NUREG/CR-3299, " Core Melt Materials Interactions Evaluation,"

Final Report (in publishing).

Enclosed are extracts of the first two monthly reports from the NRC l-Technical Assistance Contractors, R&D Associates.

l NRC staff state that additional information requested in Item 2 was supplied in response to a previous FOIA request, F0IA-83-432.

A copy of that. request and response is enclosed.

Other NUREG's which are relevant to this request are:

1.

NUREG/CR-2182 Vol. I & II

" Station Blackout at Browns Ferry Unit One-Iodine and Noble Gas Distribution and Release" (September 1982) l

Charles W. Elliott, Esquire.

2.

NUREG/CR-2572 Vol. I & II "SBLOCA Outside Containment at Browns Ferry Unit I - Accident Sequence Analysis" (November 1982) 3.

HUREG/CR-2973

" Loss of DHR Sequences at Browns Ferry -

Unit I - Accident Sequence Analysis" (May 1983)

.These reports are available by contacting the NRC Public Document Room

~(PDR),1717 H Street, NW, Washington, DC 20555.

'his 'ompletes action on your request.

T c

Since ely, i

./

. M. Felton, Director Division of Rules and Records Office of Administration

Enclosures:

As stated i

,.-c r

- ew-

MONTHLY PROJECT STATUS REPORT September 15, 1983 Report No.

RDA-MR-1273OO-003 Period Covered:

July 30 through August 31, 1983 Name of Program:

Severe Accident Mitigation Systems Contract Number:

NRC-03-83-092 Start Date:

June 27, 1983 Completion:

27 months CECTION A Technical Summary of Project Status Technical progross has been good and active work continues on Tasks 1-5.

(Task 6 is done only on specific assignments).

As agreed to at the kick-off neeting (7/12/83), the first specific containment type to be studied in Task 3 is the Mark II BWR, as exemplified by the Limerick plant.

We have been requested to make a special rapid response on our study of this type plant by January 1,

1984.

This mini-study will include assessments based on Task 1,

2, 3, and 4 activities, and thus represents the first priority of activity for all tasks.

During August substantial progress was made on reviewing Mark II containments and their failure modes, including preliminary assessment of dominant risks, possible mitigation schemes, and a preliminary value/ impact range.

Several types of mitigation systems were'solected for preliminary design and costing.

These were chosen on the basis that systems having unequivocable modes of performance should be the standard against which other systems, having greater operational uncertainty,'should be compared.

These comparisons, and the final selection of a system for design, are expected to be accomplished during October, An addendum to this monthly report provides a summary of preliminary value/ impact assesumont being performed under Task 4 for the Mark II mini-study.

This assessment provides an estimate of the range of costs that can be supported for mitigation, (including not only equipment but procedures and outage costs).

Another addendum is a draft version of the cumulative mitigation r' quirements for the Limerick plant.

e It is by no means final at this time.

The conference on LWR severe accident evaluating was attended by W.E.

Kastenberg and I.

Catton.

A meeting of SNL, NRC, Purdue and RDA was accomplished during the ccnference.

A report on the meeting will be forwarded under separate cover.

At the request of Dr.

J.

Meyer, I.

Catton attended a meeting on MF110R.

His report on the meeting will also be forwarded under separate cover.

1 3

For. Tasks 1 and 2'the process of data collection is well along,' outlines have been made of the topical reports, and some text preparation has begun.

A special list of all known mitigation proposals, devices, and systems was prepared and delivered to NRC; a copy is attached.

The report outline for Task 3 is currently being revised, and procedures and outlines for Tasks 4 and 5 are being prepared.

The Master Plan outline for the project is under way, and is expected.to be completed during the current period.

NRC has designated GESSAR as the second of the three_ types of containment to be studied in Task.3.

Plans are being made to visit the Perry plant during the, current period.

For the purposes of our contract " mitigation" has been defined as those actions, devices, components and systems that deal with events and their consequences a7'ter the core is melted.

This definition is quite straightforward except in the case of failure to scram.

For a BWR under certain conditions (Class IV) a large steam spike could rupture the containment long before core-melt occurred.

Similarly, there are some sequences (Class II) wherein the containment fails-prior to core melt, but on a long time

- scal e.

We have requested written instructions on how to

' interpret these special cases.

It is feasible to include a mitigation system for an ATWS event (probably a vent), but it.has been our impression that prevention of an ATWS would be more cost-effective than mitigation, and that the Limerick plant had already been modified to reduce such events to a non-dominant risk level (the so-called 3-A fix).

The significance of ATWS to mitigation will be clarified in the near future.

SECTION B:

Technical Status by Tasks IAEh is Sunggy gi Cgniainment Systems This task comprises z

data gathering, categori:ation of dominant accident sequences, and evaluation of mitigation opportunities f or the major types of reactor containment.

The work will be organi=ed and reported by containment type, vir. Mark II, Mark III, GESSAR, etc.

Thus the work on Mark II containment will form a chapter both in the Mark II mini-study and in the topical report for Task 1.

a.

Efforts completed:

Assessment of Mark II containments and the severe accident research literature has been completed.-

Work on the other pressure-suppression types is well underway.

Personal contact and telephone interviews have been made with research investigators in the field at 2

Sandia, Brookhaven, and Purdue, b.

Problems or delays:

None.

c.

Summary to dates Data collection mostly complete for pressure-suppression containments, well underway for other types.

Review of accident sequences is complete for BWR's, just beginning for PWR's.

Identification of mitigation opportunities is complete for Mark II containments, underway for other pressure-suppression types.

d.

Plans for next period:

Meetings have been scheduled with representatives of Sandia, Brookhaven, and Purdue to discuss and coordinate our survey with current research efforts.

It is planned to finish source collection, finish assessment of Type I and Type II ccatainments, and begin work on PWR containments.

Text preparation will also begin.

IAsk'2. Surver gi Mitigatign System 3sa. This task will survey a wide range of concepts, proposals, devices, and systems for mitigating the consequences of severe accidents.

These will be categorized into groups by function, and ranked according to f easibility, cost, etc.

The specific devices, sub-systems, etc., to accomplisn a given function, such as core-retention or heat removal, are designated as components.

A selection of appropriate components will be mede to form,a mitigation system suitable for each major containment type as determined in Task 1.

a.

Efforts completed:

Tabulation of mitigation concepts applicable to Mark II containments has been completed.

A complete list of all known suggestions for mitigation systems or components has been completed and submitted (see Addendum),

b.

Problems or delays:

None.

l c.

Summary to date:

The information collection phase of this l

task is nearly complete. Assessment and ranking of the material is under way.

An outline of the final report has been prepared and written t e:t t is being prepared.

Information'concerning mitigation components for Mark II plants has been transmitted to the other task efforts.

i l

d.

Plans for next period:

Complete collection of data, continue assessment and ranking of mitigation components.

IEEh El E25iED sod Egggibility This task comprises selection of up to three major types of reactor containment with the approval of the Project Officer, establishing for each type the requirements for a mitigation system in view

- ~

offthe. dominant risks estab1'ished in Task 1,-and choosing a

~* -

suitable combination,of components as.characterised in Task' r

~

~2.

After establishing by
preliminary _ analysis in Task 4 I

that-the functions 1 chosen to be performed would likely have a-suitable _ef f ect;on overall risk :an'd with enough design

~

effort to show probable feasibility, aEfinal selection of a.

mitigation system will,be'made,with,the approval of the Project. Officer.

Then a1 complete-conceptual design, cost i

cand feasibility assessment will be performed, a.

Efforts. completed:

Based,on. Task 1 results, the frequirements.forL mitigating the residual l risk in Mark II containments have-been established, subject to some remaining uncertainty as to whether - ATWS events are considered to have been prevented or are to be included in the mitigation category.

Prelimina'ry design-and costing has been'made'for several possible mitigation systems (combinations cf components).. These are now undergoing 4

i

. evaluation before final selection and' presentation to the

. Project Officer.

b.

Probl' ems or delays:

None.-

~

c.. Summary to_ dates For_ Mark II containment as exemplified by the Limerick Plant, mitigation requirements (functions) have been identified, including containment heat removal, core residue cap.ture and retention without concrete attack, and' (if ATWS events. are to be mitigated) some kind of venting system.

. Candidate components to fulfill these requirements have been selected for preliminary conceptual design and cost estimation._

Separate cost figures _will be generated for 1) Plants before construction _begins,

2) Plants built but not yet in cperation, and 3) Operational plants.

4 d.

Plans for next period:

Complete preliminary designs and i

_ assessments for Limerick, and begin final design of selected version.

IAEh 31 MalMRilGEaEi 6031ysis This task will provide a s

. quantitative assessment of the relative risk that can be I

averted'by mitigating particular aspects of the containment failure.

It will determine whether a proposed mitigation 4

system'is cost effective, and which components are most

- i mp or t an t. -

l a.

Efforts completed:

Collection of source documents for ji Value/ Impact evaluation is1 virtually complete.

The recent Boston meeting on: Severe Accidents-will provide addi ti on al

. material.

Copies of the mostLimportant papers are already available.

Previous Commission statements and action in'the i

l-I

~

Value/ Impact evaluation is virtual y complete.

The recent Boston meeting on Severe Accidents will provide additional material.

Copies of the most i mportant papers are already available.

. Previous Commissi'., statements and action in the field has been reviewed incluting the PNL Value/ Impact Handbook under development for RER Division.

A preliminary Value/ Impact analysis f or the Limerick plant was initiated using the BNL. review of the Limerick FRA and the new PECO PRA.

The results of this ' analysis are summarized briefly in the Addendum attached to this report.

b.

Problems or delays:

None.

c.

Summary to date:

Methodology for Value/ Impact analysis of mitigation conceptual designs is being formulated, based on prior work in the field and the specific requirements *of this task.

A preliminary analysis f or the Limerick plant has been completed.

d.

Plans for next period:

A working procedure for assessment will be put inte preliminary operation for examination of Mark II mitigation schemes.

This procedure will be discussed with NRC befcre final adoption.

Issh 51 LigEO9109 Sital22% 9EM219ESSOtz This task will assist.the NRC in utilizing.the results of.this project by developing suitable methodology and strategies for assessing and implementing severe accident mitiga, tion policies, a.

Efforts completed:

Collection and review is under way of suitable documents, reports, and prior action cf the NRC, especially SECY-82-1-A and 1-B, and NUREG 0933.

b.

Problems and delays:

None.

1 c.

Summary.1to datei Background and critoria under i'

d ev el'opinen t.'

M.

Plans.for nest period:

Continue development.

I h

IB1h si censuitatiga god Saggigi 6ssigomggtst No effort has 1,,

been assigned.by NRC to this task so far.

y,

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n 9

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f 5

+-

T

-4

ADDENDUM TO MONTHLY REPORT September 15, 1983 Value/ Impact for Limerick Mitigation In order to obtain a range of possible acceptable costs o'f mitigation'for the Limerick plant, two estimates were

.made using the proposed trial goal at $1,000/ man rem averted as follows:

ESTIMATE NO. 1

~

The first estimate uses the data contained in the BNL review (NUREG/CR-3028) of the Limerick PRA.

Note that this data does not reflect external initiators.

The Limerick PRA and the BNL review conclude that the majority of latent effects (and population dose) result from containment failure via slow overpressurination (Class I).

In Table 1,

the contribution to latent fatalities for the dominant releases are shown as well as their frequency.

The expected value is 0.174 latent deaths / year.

Of these 0.166 latent deaths /yoar come from GPREL, the Class I sequences described above.

Using a conservative conversion

_4 factor of 2x 10 latent deaths / man rem (linear dose model, no threshold) and an allowed cost of $1,000/ man rem, one obtains an upper limit cost of 33.6 million dollars.

l L

~

a ESTIMATE NO. 2 The second estimate uses the data contai'ned in the new Limerick PRA which includes external events, but not the modifications to frequency and consequences contained i n the BNL review.

The nev, PRA indicates that fires and internally initiated sequences contribute about equally to latent effects (Class I slow overpressurization).

Seismic events contributing to Class IV (ATWS events) are also important contributors to latent fatalities.

Early fatalities stil'1 continue to be dominated by Class IV events (ATWS) but seismic vessel failure also becomes an important

~

contributor.

Accidents initiated by fires fall into existing release categories.

Random vessel failure and accidents' initiated by earthquakes required new release. cat.e.gories.

The sequences which are affected by fires are QUX, OUV and OW, which increase the OPREL (Class I) release category significantly.

The seismic initiators lead to case with various combinations of vessel and cc-ainment f ailure due to

- the seismic event itself.

For those cases where the i

containment dges agt fail seismically mitigation 'is possible.

Using the data contained i n the FRA, 220 man-rem / year can be averted if containment failure due to slow i

2 l

r e

y

._y_.

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r w

overpressuriation (only) is eliminated (perf ectly).

At

$1,000/ man-rem and a 40 year life, the allowable cost is S.8 million dollars.

DISCUSSION The values estimated above (B-33 million) are ball-park estimates and are subject to change dramatically pending (a) the BNL review of the new PRA, and (b) new work on the i

Source term.

At present they are only'for use as guidance in the design of mitigation.

A target of 15 million dollars will be used_fer screening purposes.

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9 m,

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TABLE I Frequency Latent Catecory (per year) deaths 3

-2 OPREL 7.7 x 10 x

2.2 x 10 16.6 x 10

=

4

~4 R

9.8 x 10" x

2.1 x 10 20 x 10

=

-8 4

-4 R

2.1 x 10 1,8 x 10 3.7 x 10

=

2b 4

-4 R

3.2 x 10" x

1.8 x 10 0.58 x 10

=

3

-4

~

R 4.3 x 10 x

6.6 x 10 28 x 10

=

4

~

C 1.4 x 10 x

1.4 x 10

~

1.98 x 10

=

4

-8 4

~4 CY' 7.1 x 10 x

7,4 x 10 9.98 x 10

=

4

-8 4

-4 C Y" 7.1 x 10 x

1.3 x 10 9.21 x 10

=

4

-2 17.43 x 10

/yr l

4

+-

Appendum 2 CUMULATIVE' LIST OF MITIGATION REQUIREMENTS FOR MARK II The dominant severe accident sequences for the Mark II Boiling Water Reactors result in a small number of final end states wherein the containment is breached.

To consider mitigation of these end states, it is necessary to assess them in a cumulative f ashion.

Thus, if any of them result in a plant having a total electrical failure, then all mitigation schemes considered must work in this environment.

If any of them will undergo failure by, overpressure, then all systems must be able to protect against thi s - f ailure,

etc..

On the'other hand, if none,of the dominant sequences include seismic failure of the containment structure, then we can assume that it is intact for purposes of mitigation,

.and so on.

Following.are a brief list of assumptions and policies on which a mitigation design might be based, f ollowed by a tentative set of cumulative requirements for the Limerick plant, derived by assessment of its dominant failure modes.

In C.

are listed the end point conditions in the1 containment under the cumulative worst case accident sequences.

A mitigation concept that could cope with these conditions would neeessarily handle the lessor, more probable events.

A. Ground-rules and Assumptions for Mitigation Design

1. When the behavior of the core or other material is in an uncertain or debatable situation, the uncertainty will be avoided by' designing that situation out of the sequence.

If this-i's not possible, the uncertainties will be reduced to the maniumum amount possible.

2.

Passivo action will utilized wherever possible, but l

where it technically is impossible or unreasonably costly, a

fully independent and redundant source of energy will be used.

i;

)

3.

If the containment is not overpressured, it is l

assumed that the water in the suppression pool will not i

escape, although it might drain parti ally into the secondary containment building to the level of the lowest penetration.

4.

The containment will not fail to isolate, or changes will be made to insure isolation on demand.

5.

An intact containment always presents less risk than an opened one.

1 i

L L _.

s 6.

A system that segregates and confines radioactive materials into a definite, enclosed region presents less

^

human risk than one that spreads it over several regions.

B.

Cumulative List of Mitigation Requirements.

1.

If risk assessment indicates that early containment f ailure f rom an ATWS is a dominant (>1%) part of total risk, then a vent system of some type is required.

2.

Reliable, redundant cooling of the containment is required even though there is no electric power.

3.

The molten core debris must be provided an unequivocable pathway to a locati.on where it can be retained and cooled indefinitely.

C.

Assumed Initial Conditions at Time of Meltdown.

1.

All electric power has been lost, both on-site and off-site.

2. The' suppression pool has been heated by a turbine trip from full power.

If an ATWS has occurred, the pool is saturated at the vent pressure.

3.

The normal and emergency care-cooling systems are inoperative.

4.

The emergency heat removal system is inoperative.

5.

The core has boiled dry and is in the process of melting its way through the bottom of the vessel.

i 6.

More than 50 tons of molten-steel will accompany the core into the sub-vesuel area.

7.

All the circonium in the core has reacted to form hydrogen.

I i

l After further refinement and developent of these ground I

rules and conditions, they will be used to specify the i

design of specific mi ti gation systems under Task 3.

l l

~

2 i

PROPOSED SEVERE ACCIDENT MITIGATION SYSTEMS

.The -following list is fairly comprehensive as to types of proposed remedies, but does'not attempt to include every variation, modification, and repetition within each type.

No. classification was made as to feasibility, effectiveness, or cost of the proposed systems.

I.-

CORE RETENTION DEVICES 1.

Water-co5 led crucible:

A metallic container fitted with a water jacket and placed to intercept molten core material that has escaped from the reactor vessel or from the containment.

In one version the crucible is retro-fitted to an operati.ng plant by tunneling below the base-mat, and cooled by passive thermal siphons.

2.

Flooded thoria rubble bed:

A bed of refractory pebbles is placed on the floor of the reactor cavity, with water circulating through the bed.

3.

Water-cooled refractory tiles:

Similar to the pebble bed but consisting of interlocking tiles with cast-in water passages.

4.

Pebble-bed covering cooling coils:

A metallic piping system'with a pumped water supply, placed in the bottom of the reactor cavity and covered with high-density refractory pebbles.

1 5.

High-alumina cement covering cooling coils:

A cast-in-place cement liner for the reactor cavity, with imbedded cooling coils.

l 6.

Magnesium dioxide covering cooling coils:

Cooling

-coils covered with interlocking magnesia refractory brock.

l I

I I

l-l

7.

Zirconium dioxide covering cooling coils:

As above with a different refractory brick.

8.

Graphite coverine cooling coils:

Cooling coils covered with graphite or carbon brick.

Sometimes with an outer cover of steel to prevent water contact.

9.

Borax bath:

A thick layer (12 ft) of borax bricks sealed in stainless steel, covering the bottom of the reactor cavity.

10.

Heavy metal bath (lead, uranium, or copper):

Cooling coils at the bottom.of the reactor cavity, covered with a foot or so of lead bricks, or other metal.

The lead will melt and transfer heat to the coils, but remain in place since it is denser than UO 2' 11.

Iron oxide:

A layer of iron oxide over cooling coils has been proposed, with the purpose of diluting the urania to lower its viscosity and increase volume and heat transfer surface.

12.

Basalt concrete and basalt rubble beh:

Basalt 's soluble in molten urania, and the intention is to provide a dilution of the core material.

I 1.3.

Sand core retention system:

A very large mass of sand is provided below the containment building to absorb the heat of the core material and disperse it over a large volume.

1 l

14.

Iron core retention system:

A large mass of iron is provided to receive the core material and dissipate its heat.

j 15.

Flooded cavity:

Water is added to'the reactor building l

to flood the entire cavity up to the vessel and even above it, in the hope that the core material will be kept dis-persed enough to remain quenched.

l 2

L

16.

Other active cooling systems:

A number of special t

jackets and piping system in and around the reactor vessel have.been proposed, with the intention of retaining the molten core within the reactor vessel.

II.

OVERPRESSURE CONTROL FROM HYDROGEN OR HYDROGEN BURNING 1.

Oxygen exclusion:

The containment ~ is operated with an atmosphere of nitrogen or carbon dioxide, or even vacuum.

2.

Oxygen removal:

Oxygen is removed from the containment when core damage is detected, using a combustion system or chemical absorbant.

3.

Oxygen dilution:

The oxygen content of the containment is diluted below the' flammable limit with Halon gas, water fog.or mist, foams, or sprays.

4.

Igniters:

Glow plugs or spark igniters are placed throughout the containment to burn hydrogen before it reaches an explosive concentration.

5.

Fans:

Rapid mixing of the containment air is proposed to prevent local accumulation of explosive hydrogeh mixtures.

III.

OVERPRESSURE CONTROL FROM ATTACK ON CONCRETE 1.

Special concrete composition:

The reactor cavity and basemat would be made with special concrete that does not release much noncondensible gas when attacked by core debris.

l 2.

Thin basemat:

Make the central part of the basemat thin to promote rapid escape from the containment building.

IV.

OVERPRESSURE CONTROL BY VENTING THE CONTAINMENT BUILDING 1.

Non-filtered vent:

The containment is vented through a tall stack when it reaches a dangerous pressure.

4 t

l 3

f

[

=~

2' Vent'to receiver:

The containment venting is connected to'another large, closed building to provide a larger total expansion volume and greater cooling.

A companion reactor containment.has been proposed for this use.

The receiver could also be an inflatable building or balloon, kept normally empty.

3.

Vent to a condenser-filter:

A large variety of condenser filter systems have been proposed, such as sand beds, gravel beds, water-sprayed gravel beds, scrubbers, gravel / sand,

[

water pools, sand filters, charcoal filters, chemical scrubbers, all in various combinations.

i V.

OVERPRESSURE CONTROL BY CONTAINMENT HEAT REMOVAL 1.

Heat pipes:

Passive devices that absorb heat from vapor or pool space inside containment and release it externally through an evaporation-gondensation exchange with an internal fluid.

2.

Modified heat pipes:

Heat pipes hav.ing separate liquid return passages, heat pipes with-ganged penetrations, and var'iable gas-controlled heat pipes.

~

3.

Heat exchangers:

Standard cooling coils acting as

-condensers or pool coolers, with-pumped external cooling fluid.

r l

4.

Spray coolers:

Pumped sprays inside containment, com-bined with heat exchangers in the loop.

5.

Fan coolers:

Circulating fans combined with heat l

exchangers to increase thermal transfer from containment vapor space.

t 6.

Secondary suppression pool:

Provide a larger secondary

. suppression pool to increase heat capacity of system.

4

7.

More reliable residual heat removal system:

Increase the redundancy and ruggedness of the residual heat removal system.

VI.

CONTAINMENT PROTECTION AGAINST MISSILES 1.

Missile shields:

Various structures designed to pro-tect.the containment penetrations or walls against flying debris or thrashing pipes inside the containment.

4 VII.

SPECIAL CONTAINMENT STRUCTURES-1.

Underground siting:

Location of the containment vessel in an underground cavern or excavated pit, completely isolated.from the external environment.

2.

Berm shield:

Partially underground containment building, protected by a earthen wall or berm.

Sometimes a gravel bed is included for a filtered vent pathway.

3.

Double containment:

A'second strong containment building surrounding the original containment has been proposed.

4.

Strength improvements:

For improving the pressure rating of an existing containment building, wrapping with wire, adding steel ribs, etc., have been proposed.

5.

Increased volume:

Increase the free volume of the containment building on new reactors.

6.

Strengthen safety system:

Make the essential safety systems more rugged by means of armor, bunkers, and heavier construction.

VIII.

FISSION PRODUCT REMOVAL SYSTEMS 1.

Containment spray systems:

It has been proposed to decontaminate a containment building post accident with an elaborate spray system to wash down the interior with special solutions, and a treatment system to remove the contaminants from the solutions for use.

5 W

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+9r-

+-e p

w-9 y,--y.-,-

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y p

,r

,---w

--w

,--,,-t-

--w=-

2.

Gas treatment system:

Provide a special recirculating treatment system to remove fission products from the con-tainment gas volume.

IX.

PORTABLE OR ADAPTIVE RESPONSES 1.

Pumps:

Use of portable pumps, fire trucks or fire boats to add water to the containment, keep a gravel vent bed wet, or for other requirements.

2.

Earthmovers:

Use of bulldozers, etc., to build up protective shields or berms around contaminated buildings,.

prevent flood erosion, etc.

W i

R & D Associates August 24, 1983 l

1 I

i I

I l

i

(

l 6

~

j Monthly Project Status' Report 15' August, 1983 j....,

r RDA-MR-1273OO-OOl 7 Report No.-

P'eri od. Covered:

June 27 through July 30, 1983 Nameiof Program:

Severe Accident Mitigation Systems Contract Number:

NRC-03-83-092 Start *Date:

June 27, 1983 Completion:

27 months SECTION A:

Overall Summary of Project Status Technical:

In this first month of operation, active work -

has begun on Tasks 1-5.

(Task 6 is done only on specific assignments).

'At NRC request, the first specific plant type

.to be studied is the Mark II BWR, as ex empl i f i ed by the Limerich plant.

We have been requested to make a special rapid response on our study of this type plant by January 1,

1984.

Collection of data and reports is well underway, and the Limerick containment system has been visted.

Besides the work on Limerick, present effort is directed at revisions and completion of the master plan for the project.

Financial:

As of July 30, 1983, funds expended am'ounted to S 31,202 (6.5% of total).

Funds obligated amounted to

$ 58,349 (12.1% of total).

SECTION St Technical Status by Tasks IAEh 13 gucyey o[ Containment gystgm a.

Efforts completed:

s A working outline of the final topical report has been d ev el op e*d, and a substantial portion of the source documents located and' ordered.

About third are already on hand.

The significant failure modes of Type II containments have been identified, and mitigation requirements set out.

b.

Problems or delays:

None, c.

SQmmary to date:

Data collection well underway, asuenument just beginning.

d.

Plans for next period:

Finish assessment of Type II containments, complete source collection.

_Tash 2: Sucygg gi bitigatige-gystgegua. Effor~ts completed:

Tabulation of mitigation concepts applicable to Mark II containments has been completed, and collection has begun of a complete list of all known suggestions for mitigation systems or components as a special assignment.

b.

Problems or delays:

None, c.

Summary to date:

Information collection underway.

d.

Plans for next. period:

Continue collection and assessment of mitigation literature.

~

185h 21 Dgsigg ged Eggsibility:

Effnets. completed:

Preliminary assessment of feasibility has been made for several Type I.I mitigation concepts.

These are to be given 6

2-a rough costing and then subj ec ted to preliminary Value/ Impact analysis before final selection.

b.

Problems or delays:

None.

c.

Summary to date:

Several Type II mitigation concepts are undergoing evaluation; others will be added, d.

Plans f or next period:

Complete preliminary designs and assessments, and begin final design of selected version.

Iank di Valu=4Ieeest 80alysigs a.

Efforts completed Collection of source documents for Value/ Impact eval uati on

-is under way.

Previous Commission statements and action in the field has been reviewed including the PNL Value/ Impact Handbook under development f or RSR. Division.

b.

Problems or delays:

None.

c.

Summary to date:

Methodol ogy f or Val ue/ Impact analysi s of mi ti gati on conceptual designs is being formulated, based on prior work in the field and the specific requirements of this task.

d.

Plans f or next period:

A working procedure for assessment will be put into preliminary operation f or examination of Type II mitigation schemes.

This procedure will be discussed wi th NRC bef ore final adoption.

~

.i' 185h El biERO5iOg Sttgiggy ggyglggegots a.

Efforts completed:

Collection and re'iew is under way of suitable documents, v

reports, and prior action of the NRC, e s p ec i a11,y SECY 1-A and 1-B, and NUREG 0933.

b.

Problems and delays:

None.

c.

Summary to date:

Background and criteria under development.

~

Task 6:

Consul t at i on and Sagcial 8ggigameOts. No effort ansi gnrrd by NRC to thic task no far.

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UNITED sT ATES

. I, '

I,g ' h NUCLEAR REGULATORY COMMISSION

  • * * / ",

t WASHtfJGTON. O C. 20555 1

s

...../

5g v

AU72 51533

~

Mr. Steven Sholly

' Technical Research Associate Union of Concerned Scientists 1346 Connecticut Avenue, N.W.

IN RESPONSE REFER Washington, DC 20036 TO F01A-83-432

Dear Mr. Sholly:

This is in response to your letter. dated July 28, 1983, in which you requested, pursuant to the Freedom of Information Act, documents produced by Sandia National Laboratory and/or their contractors under the NRC sponsored " Severe Accident Risk Reduction Program."

Appendix A is a listing of documents responsive to your request.

These documents are being placed in the NRC Public Document Room in F0IA file folder 83-432 in your name.

This completes NRC's action on your request.

Sincerely,,

/.-

/-

...5

,' /-

/

. M. Felton, Director Division of Rules and Records Office of Administration

Enclosure:

As stated

e g.-

,~

UNION OF CONCERNED SCIENTISTS me c_eei.uu.em,e. x.w.. s. mi. washiesie, oc 20ee. on) m.5 coo 28 July 1983 Mr. J. M. Felton, Director FREEDOM OF INFORMATION Division of Rules and Records ACT REQUEST Of fice of Administration fCI A V.3 U.S. Riclear Regulatory Commission b e g f _. /

p 3 Washington, D.C.

20555

Dear Mr. Felton:

Pursuant to the Freedan of Information Act, please make available for public inspection and copying at the ccmnission's 1717 H Street Public Docunent Roan copies of docunents in the following categories:

, A.

All documents produced by Sandia National Laboratories and/or their contractors under the NRC-sponsored " Severe Accident Risk Reduction Frogram" (SA9R).

This request specifically inc1u3es 3

draft rel orts, papers prepared for presentation at technical society meetings (e.g., ANS/ ENS meetings), and menoranda ccTmunicating results anad conclusions of this work to the NRC.

This request also specifically includes docunents concerning value-impact and/or cost-benefit analyses of risk reduction measures analyzed in the SARR Program.

I reccgnize that this request may involve a large nunber of reports.

At present, however, I am unable to refine the request further due to a lack of information on the NRC contract nunber, the identities of the Sandia researchers performing the work, or the identity of the NRC Staf f Technical ttanitor for the project.

I am, however, willing to discuss this request wi th l

your staff to help avoid unnecessary docunent search ef forts.

Should you or your staff have any questions regarding this request, please do not hesitate to contact me at UCS's Washington, D.C., of fice at 296-5600.

Your cooperation in responding to this request is appreciated.

i Sincerely,

.C.

Steven C. Sholly Technical Research Ass.' ate l

Mai. Office: 26 Church Street. Cambridge. Massachusetts 02238. (617) 547 5552

~

APPENDIX A 1.

Evaluation of Severe Accident Safety System Value Based on Averted Financial Risks, SAND 83-0443C.

2.

Evaluation of the Sensitivity of Reactor Risks to Uncertainties, SAND 83-0855C.

3. -Severe Accident Risk Reduction Program, SAND 82-2141C.

4.

Ltr:

Benjamin to Cunningham, August 30, 1982, w/following attachments:

a.

Value - Impact Analysis of Severe Accident Prevention and Mitigation Systems, 8/82.

b.

Risk Reduction Analysis of Severe Accident Prevention and Mitigation Systems, SAND 82-1697C.

c.

Benjamin slides, Sept. 2, 1982.

5.

Benjamin slides, LA ANS Mtg., June 9,1982.

6.

Ltr: Benjamin to J. B. Van Erp. Januar,y 29, 1983.

7.

Risk and Systems Interaction Analysis of Severe Accident s

Prevention and Mitigation Systems, SANDC2-0400A.

8.

Ltr:

Benjamin to E. N. Cramer, January 12, 1982.

9.

Note:

Hatch to Kolaczkowski, July 25, 1983, ASEP products expected by the SARR Program in FY83 and Beyond.

10.

Ltr:

Griesmeyer to distribution, July 22, 1983, MARCH Screening Sensitivity study.

11.

General SARRP viewgraphs, 7/83.

12.

SARRP Approach to Risk Benchmarking, May 19, 1983.

13.

Ltr:

Benjamin to Cunningham, April 20, 1983, SARRP Plan for the Accomplishment of Phase I objectives.

14.

Presentations; NRC Management Review of Sandia Severe Accident Programs.

15.

SARRP Phase I Report Outline, 9/82.

16.

Design Considerations for Implementing a Vent - Filter System at the Barseback Nuclear Power Plant, Johansson et. al., August 1982, Chicago, ANS meeting.

Re:

/01A-83-432 p

APPENDIX A 17.

Ltr:

Benjamin to Cunningham, May 19, 1982, Severe Accident Uncertainty Analysis.

18.

Presentation:

NRC review of SARRP, etc., March 19, 1982.

19.

Ltr:

Benjamin to P. B. Bleiweis, January 29, 1982, review and comments on Bleiweis ceport.

20.

SARRP presentation by A. S. Benjamin, October 21, 1981.

21. ' Cost-Benefit considerations for Filtered-Vent Containment Systems, 17th DOE Nuclear Air Cleaning Conference.

22.

Risk Assessment of Filtered-Vented Containment Options for a BWR Mark III containment, SAND 82-0403C.

23.

Probabilistic Risk Assessment of Filtered-Vent Containment Systems:

Mark I BWR, Abstract from ANS Transactions, 1981 summer meeting.

24.

Presentation by A. S. Benjamin on PRA of FVCS for Mark I BWR,1981 ANS summer meeting.

25.

Ltr:

Benjamin to N. J. Diaz, circa November 1980, summary of paper for 1981 ANS summer meeting.

26.

Filtered-Vent Containment Systems, IAEA-CN-39/103.

27.

PresentationonFVbS,A.S. Benjamin, April 19, 1982.

28.

FVCS Program - Grand Gulf Results, F. T. Harper, April 19, 1982.

29.

Ltr:

A. L. Carter, Holmes and Narver to Benjamin, January 13, 1982.

30.

Ltr:

Benjamin to M. W. First, Harvard, July 20, 1981.

31.

Ltr:

Benjamin to Rowsome, March 12, 1981.

32.

Ltr:

A. L. Carter to H. C. Walling, SNL, January 27, 1981, Cost Estimate of Toroidal Suppression Pool.

33.. FVCS-BWR Mark I presentation by Benjamin, circa 1981-1982.

34.

Issues affecting the Feasibility and Effectiveness of Vent-Filtered Containments, SANS79-ll39A.

.., e APPENDIX A

.35.

Program Plan for the Investigation of Vent-Filtered Containment Conceptual Designs for Light Water Reactors, NUREG/CR-1029, October,1979.

36.

Effect of Containment Venting on the Risk of LWR Meltdown Accidents, NUREG/CR-0138, June, 1978.

37.

A Value-Impact Assessment of Alternate Containment Concepts, NUREG/CR-0165, June, 1978.

38.

Value-Impact Comparison of Alternate Containment Designs, SAND 77-1103C, November, 1977.

a l

i l

_