ML20045D116

Summary of ACRS Subcommittee on Severe Accidents 930318 Meeting in Bethesda,Md Re Resolution of Severe Accident Issues Associated W/Design Certification Review of Ge/Abwr

ML20045D116
Person / Time
Site:	05200001
Issue date:	04/10/1993
From:	Kress T Advisory Committee on Reactor Safeguards
To:	Advisory Committee on Reactor Safeguards
References
ACRS-2873, NUDOCS 9306250348
Download: ML20045D116 (48)

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CERTIFIED BY T. KRESS 4/10/93 i

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SUMMARY

/ MINUTES OF THE ACRS SUBCOMMITTEE MEETING i

l ON SEVERE ACCIDENTS i

l MARCH 18, 1993 BETHESDA, MARYLAND

]

PURPOSE The ACRS Subcommittee on Severe Accidents held a meeting on March 18, 1993. The purpose of this meeting was to discuss the resolution of severe accident issues associated with the design certification review of the Advanced Boiling Water Reactor (ABWR). A copy of the l

meeting agenda and selected slides,f rom the presentations are l

attached. The meeting began at 8:30 am and adjourned at 3:45 pm and was held entirely in open session. No written comments or requests for time to make oral statements were received from members of the i

public. The principal attendees were as follows:

i l

l ATTEND _REE AGRS NRC STAFF T.

Kress, Chairman A. Thadani, NRR P. Davis, Member J. Wilson, NRR W.

Lindblad, Member J. Monninger, NRR C. Michelson, Member R.

Barrett, NRR R.

Seale, Member R.

Palla, NRR W.

Kerr, Consultant C. Poslusny, NRR D. Ward, Consultant J. Kudrick, NRR D.

Houston, Cognizant Staff G. Kelly, NRR Engineer C. Tinkler, RES GENERAL ELECTRIC C. Sawyer C. Buchholz J. Quirk A. Beard J. Gabor, Consultant Also in attendance were representatives of

DOE, TENERA, Westinghouse, B&W and Bechtel.

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' t Sev Acc Minutes March 18, 1993 DISCUSSION Chaiman's Ooenina Remarks In his opening remarks, Dr.

Kress discussed the Commission's guidance given in the Severe Accident Policy Statement and 10CFR52 for advanced or new plant designs. In these documents, an applicant is required to show that the proposed plant design is based on consideration of severe accident issues and that the resolution of these issues can strike a balance between accident prevention and consequence mitigation. He indicated that the major severe accident issues are associated with hydrogen generation, steam explosions

[ fuel-coolant interactions (FCI)],

direct containment heating (DCH), core-concrete interactions (CCI) and containment bypass.

History of Closure Process - A. Thadani, NRC/NRR Dr. Thadani discussed the rationale for addressing severe accidents in advanced plant designs and the key steps in 1.1 corporation of severe accidents in the review process. He also discussed the role of the design basis accident analysis and the role of the severe accident analysis, indicating the purpose, general guidelines and design guidance for each category. He provided a chronology of the resolution of issues. In summary, he stated that the resolution path for the ABWR was well established and that the staff would present the results of their deterministic evaluation later in the meeting.

ABWR Accident Prevention - C. Sawyer, GE Mr. Sawyer discussed the ABWR design features in regard to reducing the challenges from transients, LOCA, loss of heat removal, station blackout, operators, ATWS, and shutdown. He further discussed the design features that mitigate external threats (fire, flood and seismic), and the defense against common-cause failures (functional diversity and physical separation). He indicated the nature of design changes that had been made as a result of PRA studies. He presented a summary of the core damage frequency (CDF) values for type of event. The total CDF was calculated to be 1.6E-07/RY with i

a 76% contribution by loss of of fsite power and 23% by transients.

In his conclusions, he indicated that the CDF for the ABWR was very low and that both the NRC and ALWR goals for CDP were met by a large margin.

Resolution of Severe Accident Issues for ABWR - C. Buchholz, GE Ms. Buchholz discussed the ABWR design features that are important for accident mitigation. These features included the following:

pressdre suppression containment, containment isolation, drywell-wetwell vacuum breakers, reliable vessel depressurization, nitrogen inerted atmosphere, lower drywell configuration, firewater addition system, lower drywell flooder and containment overpressure w

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1 I

Sev Acc Minutes March 18, 1993 protection system. She discussed the operation and benefits of the following systems: AC independent water addition, lower drywell flooder and containment overpressure protection.

She further discussed the ABWR design features versus the accident mitigation challenges in SECY-90-016. These challenges are: hydrogen control, DCH, FCI, suppression pool bypass, CCI and overall containment performance. She indicated that the conditional containment failure probability (CCFP) was calculated to be 0.002 for large release considerations and 0.005 for loss of pressure boundary integrity.

In conclusion, she indicated that the ABWR meets the established goals with a wide margin.

Staff Evaluation and Conclusions - R. Barrett, J. Monninger, NRC/NRR Mr. Monninger discussed the status of the staff's evaluation for each of the severe accident issues that have been listed in SECY-90-016 and the follow-on documents. The following issues for the proposed ABWR design are considered acceptable or closed by the staff: ATWS, fire protection, hydrogen generation and control, DCH, containment vent design, FCI (steam explosions), and suppression pool bypass.

The issue of station blackout is classified as confirmatory. The following issues are still being pursued as open or under evaluation: intersystem LOCA, containment performance, CCI, and lower drywe]1 floor sump design.

In concluding remarks, Mr. Barrett stated that the GE ABWR is a good example of a design that addresses the severe accident issues.

He indicated that the staff would continue to pursue 'those issues that are still open and keep the ACRS informed about their resolution. He acknowledged the homework assignments that had resulted from the discussion. These are characterized as follows:

(1) RWCU rupture scenario as it affects the HVAC system and common cause failure of ESFs, (2) rereview of the ACRS report on hydrogen control as it relates to stratification, (3) pH control in the suppression pool in regard to revolatization of cesium iodide, and (4) locate the comprehensive regulatory review of the containment overpressure protection system.

Subcommittee Comments. Concerns and Recuests During the meeting, Subcommittee Members and Consultants expressed various comments and concerns as follows (random order) :

(1) Mr. Michelson expressed a concern that the scenario of containment bypass was not being addressed adequately. Dr.

Thadani agreed to have the staff look at it further. Mr.

Palla (NRR) indicated that the ABWR PRA studies showed this to be a very low probability event, in the order of 1E-09/RY.

i Sev Acc Minutes March 18, 1993 i

i (2) In regard to flood protection, Mr. Michelson expressed a i

concern that large pipe breaks outside containment were not being addressed.

(3) In regard to physical separation of systems, Mr. Michelson 4

expressed a concern about the separation criteria for HVAC systems.

l l

(4) Regarding the staff's acceptance of a 10% limit for hydrogen concentrations for the explosive level, Mr. Ward indicated that past ACRS reports on this issue did not agree with this finding. The staff agreed to take another look at the ACRS reports on hydrogen and take the Committee comments into l

consideration.

(5) Also, in regard to hydrogen generation and control, Mr.

Michelson expressed a concern about possible hydrogen explosions either in the vent pipe or in the reactor building when this gas was discharged into those areas.

i (6) Mr. Ward questioned the double fixation proposed for DCH resolution, i.e.,

both reliable depressurization and cavity design. He indicated that the ACRS had recommended that only one of these mitigation features be provided. The staff stated that GE had proposed both in the design.

(7) In regard to the overpressure protection system, Dr. Kerr noted that the staf f had approved the resolution of the issue subject to the results of a comprehensive regulatory review. He asked where this review report could be found since the staff had closed this issue. The staff agreed to look for the report.

(8) In regard to revolatilization of iodine, Dr. Kress commented i

that he thought the wetwell would go acidic in less than 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, thus giving rise to the evolution of iodine gas.

Ms. Buchholz indicated that they think it will stay basic due to cesium hydroxide concentrations. The staff agreed to take another look at this matter.

(9) In regard to the deterministic containment performance goal based on ASME service level C limits and an integrity period of at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, Mr. Ward indicated that this was more conservative than the usual value of 0.1 for CCFP.

(10) In regard to a question about the availability of the ABWR PRA, GE indicated that the final version would be' submitted in the July-August, 1993 time frame. This would include an a~nalysis of both internal and external event and would compare the ABWR with the NRC Safety Goals as well as the EPRI Utility Requirements Document (URD).

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Sev Acc Minutes March 18, 1993 Closino Remarks In

closing, Dr.

Kress thanked GE and the staff for their participation in the meeting and noted that additional meetings would have to be held to complete the review of this matter.

FUTURE ACRS ACTION The Subcommittee will schedule future meetings on this matter as appropriate. The NRR staf f has proposed an intensive one month ACRS effort on the review of the ABWR after GE submits the final SAR.

ACTIONS. AGREEMENTS AND COMMITMENTS The following actions, agreements and commitments resulted from this meeting:

(1) The staff agreed to study the following: (a) RWCU rupture scenario as it affects the HVAC system and common cause failure of ESFs, (b) rereview of the ACRS report on hydrogen as it relates to stratification, (c) pH control in the suppression pool in regard to revolatization of cesium iodide, and (d) locate the comprehensive regulatory review of the containment overpressure protection system.

DOCUMENTS The review documents for this subcommittee meeting were as follows:

(1) Advanced Boiling Water Reactor (ABWR) Standard Safety Analysis Report, Chapter 19, Response To Severe Accident Policy Statement (Thru Amendment 22), Proprietary Information (2) Memorandum for Dennis Crutchfield (NRR) from Ashok Thadani (NRR) dated September 18, 1992.

Subject:

GE ABWR Severe Accident Closure Chapter (Draft FSER)

NOTE:

Additional meeting details can be obtained from a transcript of this meeting available in the NRC Public, Document Room, 2120 L Street, N.W.,

Washington, D.C.

20006, (202) 634-3273, or can be purchased from Ann Riley & Associates, 1612 K Street, N.W.,

Suite 300, Washington, D.C.

20006, (202) 293-3950.

4 i

s WILLIAM KERR 2009 Hall Ave.

Tel 313482 8701 Ann Arter, MI 44104 Faz 313 763-4340 wdliamforr@um.cc.unuch.edu 23 March, 19,93 Mr. Dean Houston ACRS

Dear Dean:

Below are my comments on the meeting of the succommittee on Severe Accidents of the ACRS held in Bethesda on March 18, 1993.

COMMENTS ON SEVERE ACCIDENTS SUBCOMMITTEE MEETING

1. The presentations made by the GE Company's representatives were well organized and appeared to be responsive to the staff's concerns.

The GE group investigating severe accident phenomena is making considerable use of the MAAP code.

The resulting conclusions, are heavily dependent upon the models, the assumptions, and the limitations inherent in this code.

They recognize the uncertainly of their results, and appear to depend upon sensitivity analyses to compensate for the uncertainty.

Sensitivity analyses have a limited capability to compensate for code limitations.

In its final review of severe accidents the committee should ask the staff to orovide the bases for its accentance of those of GE's conclusions that decend on the results of MAAP based calculations.

2. GE has made a number of changes to decrease core damage frequency (CDF) and has also maos changes in design to mitigate the consequences of severe core damage.

Among those meant to decrease CDF are the automation of a number of emergency procedures which previously required operator action.

Implicit in this approach is an assumption that the resulting system will reduce the CDF.

However in order to achieve automation, a number of systems have been added.

These add to the complexity of the totar reactor system, they will require periodic maintenance and testing, in the event of failure of these systems the operators are likely to be confused.

Neither GE nor the staff commented on whether analyses have demonstrated that the added systems produce a decrease in risk.

In the final review this cuestion should be investicated further.

3. The presentation of the staff indicates that their principal emphasis in severe accident mitigation is on containment performance.

It appears that the only significant change in required containment performance, since the time that severe accidents came to be recognized by the staff and the n

2 i

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i Page Two Houston j.

d Commission, is a change in the amount of hydrogen production with which-a containment system must cope'without failure, and a requirement for a minimum design pressure,. independent of the j

containment type._

There is'no requirement on containment temperature, e.

g., even though containment components such asi j

penetrations, may fail _at temperatures'that are-likely to be reached during_the course of.some of the postulated. severe accidents. This cuestion deserves'further= consideration.

4.

Both the staff and GE apparently use the results of the PRA i

being performed by GE to-discard certain postulated' sequences _

j associated with severe accidents (e.g. ex-vessel steam-l explosions) on the~ basis of low probability.

It-is crudent to i

use not only the erobability. but to consider the risk as well.

when decidina whether to discard a seauence. This-is-especially true for those sequences for which the uncertainty in calculated risk is large.

i

5. Accident mans;2 ment is apparently;to be the responsibility of

]

the COL.

Although some of the details of organization'and 1;

procedure for accident management:may depend on the-individual j

owner-operator, it appears that for a standard plant:there.

i should be less diffa.nce1among plants than for the current j

generation of custom plants.

Since many of the crocedures associated with accident manaamment decend on details of clant desian, it annears that considerable auidanca in this area

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should-be develooed by the olant-desianggt 1

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6. The staff indicated that.there are calculationsLnow:undarway, j

l' using CORCON, to evaluate pressure build up in_the containment

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during the course of a severe accident.

A key element in the

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calculation will be the amount, timing and state of the corium entering the containment.-

The GE presentation indicated that j

those making their calculations assumed that an amount of.

j corium equivalent' to about 10 per cent of the. original core was

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used in their calculations.

(See, however,. Appendix G in j

NUREG/CR 5809 which suggests that a number larger than 10 per l

j cent is appropriate.)

This issue-deserves further consideration.

.j

7. The separation of the review of Severe Accident analysis from 7-that of the PIUL appears artificial, especially in light of the use of.the results of the PRA in deciding'to exclude certain severe accident sequences from further consideration.

Ihm committee should bear this in mind when reviewina these two

]

l areas.

It is to be hoped that the designers considered these j

two analyses as fully integrated.

4

8. GE has made a good faith effort to treat the Severe Accident Issue, with-limited guidance.from the staff.

The staff-conclusion might be described as, "We are convinced that.

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this design will result in a plant with less risk than existing-BWRs.

We aren't sure how much, and we don't propose to try to calculate how much, but we believe that GE has done enough."

Since this review will set a crecedent for review of any other 3,volutionary niant, the Committee should investicate what the staff has learned from this review, and what, if any chances in-anoroach or success criteria it exnects to make in review of evolutionary desians yet to be reviewed.

9. During the Staff's presentation there was a transparency (#4 of

)

Dr.-Thadani's-presentation) that purported to describe the process that will' be used to obtain " closure" of severe accident issues.

I reco==and that the Committee continue to exnlore how-the staff will determine when closure has occurred.

At one extreme one might conclude that going through the process described by_ the transparency, no matter what the results, would achieve closure.

I doubt that this was intended.

It would be helpful-to the ACRS review to know what is intended.

I hope these comments are helpful to-the Committee.

Sincerely, M

William Kerr i

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/cc Dean Houston l

l April 10,1993 i

To:

Tom Kress, l

Chairman ACRS Severe Accidents subcommittee i

i From: Dave Ward, i

ACRS consultant a

3

Subject:

Comments on meeting of March 18 on review of ABWR design and severe accident considerabons l

You asked to be provided comments on what ACRS should hear next on the subject.

I suggest it would be useful for the Committee to leam more about the following:

1 ATWS G]E opted to incorporate a diverse scram system rather than credit, through analys i

possible benign system response to an ATWS. ACRS had suggested that the latter would be acceptable and the Staff and Commission had agreed.

did GE opt for j

the added system? Was the calculated system response not benign 2 Unnecessary requirements A] number of features are being included in the ABWR des n, even though they not be warranted by severe accident concerns, because G says they are easily -

4 accommodated. This may establish a precedent that these things are actually i

necessary in other future designs or in possible future ABWR modifications. These indude:

GE said the hydrogen recombiners " required" by RG 1.97 will cost "only a few 1

million dollars". Recombiners do no for severe accidents. Is there really a need for them? Good DBAs should su for the things that really might

[

happen. What does RG 1.97 have in mind? A million here, a million there; pretty i

i soon it adds up to real money - to paraphrase Everett Dirksen.]

i Why are both the belt and suspenders of venting and cavity design required for j

l the rare events lead to DCH. The Staff DCH is a risk-dominant sequence.

is that true for ABW Would it be with on the belt?

j PCT > 2200 F as a definition of core damage appears to be unnecessaril conservatrve. Apparently GE can tolerate this, but is it a good precedent? y j

31 The so<: abed deterministic cenLikwnent reauirement i

IHow does the determirusDc criterion for containment, " Service Level C", etc. have anythiry to about uncertainhos in, e.g., phenomenological behavior, as the Staff clamec on p.

ofits presentabon charts?

4' Hvdrooen detonation and stratificabon

'he Staffclams.20 of charts] that using 10% rather than 13% as a detonation a

threshold sufficient mar for concems about high local concentrations of a

hydrogen, especially as caused strabfication. This should be explained and justified.

5] MAAE Some examples of MAAP calculations, perhaps compared with anything MELCOR has done, should be shared with ACRS. Especial ofinterest would be calculations which influenced decisions about whether an AB feature was acceptable or not.

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l 6] Stratification and hoh temperatures One of the items suggested for consideration in the ACRS letter on May 1991 were possible detnments to containment integrity that might result from local temperature extremes due to stratificabon of hot gases in an accident, perha j

This seems not to have been considered in the ABWR review. ps even in a LOCA.

The Staff should l

explain why they consider it unnecessary.

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7 IS-LOCA-l T w intersystem LOCA issue is still open. ACRS should hear about its resolution.

8] How much metalis available?

GE said that 100% of dadding in contact with the fuel amounts to about 30% of the total.

I metal in the core area. This 100% is often presented as if it is a conservative upper limit.

j If there is some reason to believe the remaming 70% is categorically less likely to produce hydrogen, GE or the Staff should expain that. Otherwise, they should justify use of the 30% value.

9 MCCI and 0.02 so. m) MWt This is not resolved and ACRS should hear more.

On other subjects:

It seems to me that the steam explosion issue for ABWR has been adequately resolved.

+The fact that NRR has combined severe accident and containment issues into one branch shows real progress. For too long, the Staff behaved as if these were almost unrelated subjects. They havent yet gone the full distance to a rational perspective, however. Charts 5 and 6 retain some muddle in the PURPOSE sections, although the -

i uidelines and requirements are not bad. SECY-92-070 still maintains that SA and l BA "should be kept separate", if this means separate in the sense that design requirements are different, as summanzed on charts 5 and 6, that is good.. But, if they retain the idea that SA should not be a basis for designs, that is not good.

n 9

O N

f ACRS SUB-COMMITTEE MEETING SEVERE ACCIDENT CLOSURE FOR THE GE ABWR MARCH 18, -1993 W

Ashok Thadani, Director N

Division of Systems Safety and Analysis, NRR Richard Barrett Jack Kudrick' John Monninger Mike Snodderly Containment Systems and Severe Accident Branch, NRR 9

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- Provide rationale for addressing ' severe accidents in advanced i

designs i

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- Define the closure path for severe accidents for the ABWR

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- Define the role of severe accident evaluation in the ABWR l

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RATIONALE FOR ADDRESSING SEVERE ACCIDENTS Commission Policy: Advanced nuclear power plants must address severe accidents in the design certification process

- Severe Accident Policy Statement (1985)

Compliance with current regulations including 10 CFR 50.34(f)

Resolution of USis and 1.iedium and high priority GSis Completion of a PRA Staff deterministic engineering analysis complemented by PRA

- Safety Goal Policy Statement (1986)

Two qualitative safety goals supported by quantitative objectives Emphasize features such as containment and defense in depth Proposed the general performance guideline that the overall frequency of a large release would be less than 1E-06/RY

- Standardization Policy Statement (1987)

Applicants for design certification should address the four licensing criteria in the Severe Accident Policy Statement

- 10 CFR Part 52 (1989)

Codified first three criteria of the Severe Accident Policy Statement

- SECY-90 016 (1990)

Outlined the basis of the staffs deterministic engineering analysis 3

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1 KEY STEPS IN INCORPORATION OF SEVERE ACCIDENTS Closure of severe accident issues is obtainable during the design l

certification process:

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- Addressing severe accidents in the design process l

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-Incorporating design features through lessons learned Severe accident research Operating experience Analytical insights Probabilistic Risk Assessments l

- Assessing additional significant and practical improvements for severe accident prevention and mitigation (10 CFR 50.34(f)(1)(i) and NEPA review for SAMDAs) l

- Developing an Accident Management Plan I

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ROLE OF DESIGN BASIS ACCIDENT ANALYSIS PURPOSE:

- To ensure an adequate level of protection

- To prevent core damage

- To limit offsite releases

- To meet federal requirements GENERAL GUIDELINES:

- Based on non-mechanistic event

- Use of conservative analysis

- Supported by methodology provided in Chapter 15

- Spectrum of pipe breaks and locations considered i

DESIGN REQUIREMENTS:

- Single failure criteria

- Safety related systems

- Conservatism addresses uncertainties

- Environmental qualification 5

4

ROLE OF SEVERE ACCIDENT ANALYSIS PURPOSE:

- To achieve an increased level of safety

- To balance accident prevention and consequence mitigation

- To investigate the margin provide in the containment design j

- Basis for accident management program GENERAL GUIDELINES:

- Realistic event scenarios

- Best estimate analysis

- Engineering judgement

- Use of computer codes with recognition of uncertainty (MAAP, MELCOR etc.)

DESIGN GUIDANCE:

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- Single train

- Commercial grade

- Equipment survivability l

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SUMMARY

- Advanced reactors must address severe accidents in the design process

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- Resolution path for the ABWR is well established

- SECY-90-016 provides basis for deterministic evaluation

- GE will present an overview of their design l

- Staff will present-the results of the deterministic evaluation e

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GENuclearEnergy 1

ABWRAccidentPrevention Summary

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Presentation to Advisory Committee on ReactorSafeguards i

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C. D. Sawyer, Manager,

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ABWREngineering i

March 18,1993 i

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1 ABWR Design Features

• Transientchallengesreduced

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Wide range sinatup monitoring system with period-basedprotection

- Redundant coneelsystems win self-test & self-diagnosis

- Full 2/4 trip logic wie self-test & self-diagnosis

- Three 50% Feedwaterpumps; auto runback of RIPspeedifP pump trips

- 10Reactorintemalpumps

- Power-actuatedSafety/ Relief valves i

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- Auto power runbeck on reduction of feedwater temperature

- Auto runback of RIP speed on loss of a circ waterpany 1

- AdvancedMMI,hection-oriented

- Semi-automation ofroutine tasks l

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ABWR Design Features (Cont'd) o

• LOCA challenges reduced I~

. Elimination of RPVlarge piping below core elevation division ECCS with high pressure and low pressure pumps in each division

- Increased rating oflowpressure piping

- RHfCU drainlineisolation valve I

• Loss of heat removal challenges reduced

- 3 divisions of heat removal in all RHR modes

- RHfCUavailable athighpressure

- Centainmentoverpressure relief I

• Station Blackout challenges reduced

- 2offsitepowersources

- 3onsite dieselgenerators 1

- Combustion gas turbine (auto start and connect to investment protection loads) l

- RCIC(steam driven) 3

- Firewstercross-tie to RHR i

1 ABWRDesign Features (Cont'd)

• Operatorchallengesreduced

-- Automatedresponse to most events n Auto trip of 1 feedwaterpump after scram to avoid oss of whole system from high waterleveltrip a Separation of RCIC and HPCFinitiation levels n ADS dryweIIpressure bypass timer j

n Suppressionpoolcooling i

n Post-LOCA heatremoval 1

xATWS j

- AntemetedguidanceofE0Ps

- Built-in manuallogic bypesses to facilitate execution of E0Ps (replacesjumpering)

- SigniHcantreduction ofnuisance alarms

- SPDS integrated into the normel display system e

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ABWRDesign Features (Cont'd) 4 e

• ATWSchallengesreduced

- Acennaulator-driven scram withoutScrem Discharge Volume

- EntMD electic nm-in l

- AlternsteRodinsertion i

- RPT of 4 pumps, followed by RIP speed runback

- Autohredwaterrunback

- Auto ADSinhibit

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- AutoSLCSinjecdon

• Shutdownchallengesreduced

- 3dedicatedRHRshutdownloops

- 2Owins ofFuelPoolCooling i

- All core cooling pumps potendally available (except RCIC)

- Large waterinventory over hool-long does required to reach boilingpoint l

ABWR Design Features that Mitigate External Threats

• Hre Protecdon

- Separadenamong divisions

- Hrebarriers

- Hreprotectionsystem

- HVACin smoke removalmode

- Dedicated cooling for safety-grade equipment

• RoodProtection

- Separationamongdivisions

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- Use of water-6ght doors and curbs limits aNected areas

- Autopump trip andisoladon for service water, circulating water

- Use of vacuum breakers ifsite warrants it

• SeismicPtetecdon

- Designedhar0.3g all-soils

- Impmved supports for RHR heat exchangers i

- High conHdence lowprobability failure >0.6g

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ABWR Defense Against Common-Cause Failures

• Func6enaldinasily

- Conkelreds/boren

- Reactorprotecdon systenValternate rodinsertion

- Highpressure/lowpresseneinjection o

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- Meter) turbine \\fdieseldrivenpernps

- Dieselfgasharbinegenerators 4

- ShutdomVsuppressionpoolcooling

- Digital /hardwiredprotection l

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• Physicalseparation

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- Electicaldivisions

- Redundantkeins i

- Diversesystennifequipenent

- ConealroonVRemoteshutdownpanels

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ABWR Design Changes Made as a Result of PRA Studies i

• 3-divisionECCS

• ACindependentwateraddition

• Gascombustionturbine

• RWCU Riter domin bypass to permit decay heat removal at high I

pressure l

• Conteimnentoverpressureprotection

• Automation of Suppression pool cooling

• AutomadonofATWSmitigation l

• ADSdrywellpressure bypass timer l

• RCICcapabilityforlocalcontrol i

• 9 SRVaddedto Remote Shutdown Panel i

• Increasedrating oflowpressure piping to eliminate ISLOCA concerns l

• RWCUdrainlineisolation valve

• Service Waterand Circ Waterpump trips andisolation on flood i

• Improved RHR heat exchanger supports i

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SummaryofResults

'CDFby Type ofEvent 1

j Type ofEvent CDF

% of Total i

• Transients 3.7E-N D

• Loss ofoWsitepower 1.2E-07 16

• LOCAs 6.9E-10

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• ATWS 2.7E-10

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AFTERNOON BRIEFING STAFF EVALUATION AND CONCLUSIONS The Severe Accident Policy Statement prescribed a deterministic j

engineering analysis of severe accident risks and judgement cornplemented by PRA Basis for the staff review and acceptance criteria was provided in SECY-90-016 and follow-on papers Several severe accident challenges were addressed through prescription of oesign criteria in SECY-90-016 Anticipated Transients Without Scram Station Blackout Fire Protection Intersystem LOCA Hydrogen Generation and Control Core-Concrete interaction High Pressure Core Melt Ejection ABWR Containment Vent Design l

SECY-90-016 also provided guidance for evaluation of containment performance Containment Performance Goal i

ASME Service. Level C Containment Ultimate Pressure Capability Several severe accident issues were evaluated to determine their impact on containment performance l

Core-Concrete Interaction Fuel Coolant Interaction Suppression Pool Bypass Lower Drywell Floor Sump Design 12 i

e lNTERSYSTEM LOCA PURPOSE: To reduce the possibility of a LOCA outside of containment SECY-90-016: - To the extent practical, design systems connected to the RCS to an ultimate rupture strength at least equal to full RCS pressure

- Systems not designed to full pressure should provide capability for leak testing of isolation valves, valve position indication, and high pressure alarms ACRS LETTER:

- Recommended approval of staff position provided consideration is given to all elements of the low pressure piping system (e.g., instrument !ines, pump seals, heat exchanger tubes, and valve bonnets)

STAFF RESPONSE: - Agreed with the ACRS comment DRAFT CP: - Position unchanged COMMISSION SRM: - Approved the staff's position as supplemented by the ACRS comments GE PROPOSAL:

- GE providing system by system evaluations and upgrades STAFF EVALUATION: - Staff believes systems should be designed to at least 0.4 times design pressure (410 psig)

- Staff conclusion pending further review of GE submittals STATUS: - This issue is open 18

HYDROGEN GENERATION AND CONTROL PURPOSE: - To provide a system for hydrogen control that can safely accommodate hydrogen generated by the equivalent of a 100 percent fuel-clad metal-water reaction

- To preclude uniform concentrations of hydrogen from exceeding 10 percent by volume or provide an inerted atmosphere SECY-90-016:

- Requirements of 10 CFR 50.34(f)(2)(ix) remain unchanged ACRS LETTER: - Support the staff's recommendation for amount of hydrogen to be considered

- Suggest that the staff seek further technical information on possible effects, including stratification, before establishing a limit for the average hydrogen concentration STAFF RESPONSE:- An analyses of detonation loads that would threaten containment would be extremely complex and coupled with large degrees of uncertainty

- Staff actions are to assure that the likelihood of both global and local detonations are minimized

- A review of available detonation test results indicates that 10% is a reasonable limit to assure non-detonable mixtures

- True limit lies between 10% and 13%

- 10% criterion provides a degree of margin to address the issue of local detonations and the potential for stratification resulting in concentration gradients COMMISSION SRM: - Approved the staff's position

- Staff should seek additional technical information and advise the Commission if reconsideration is warranted DRAFT CP: - Position unchanged 1.9

GE PROPOSAL: - Nitrogen-inerted containment atmosphere Hydrogen burning and explosions are not possible

- < 1 % of plant operation time, containment is not inerted

- Hydrogen recombiner system for design basis accidents

- Analysis provided in 19E.2.3.2 for 100% MWR

- peak pressure of about 75 psig

- Service level C is 97 psig STAFF EVALUATION: - 100 percent fuel-clad metal-water reaction is a surrogate for both in-vessel and ex-vessel hydrogen generation

- 10 percent uniform concentration was established to prevent detonable mixtures

- Loads from hydrogen generation from the metal-water reaction are limited to pressurization from the hydrogen gas produced and the additional heat from the exothermic reaction

- Peak pressure associated with 100 percent metal water reaction is less than Service Level C

- ABWR meets criteria of 10 CFR 50.34(f)

STATUS: This issue is closed 20

HIGH PRESSURE CORF MELT EJECTION r

PURPOSE: - Decrease tha likelihood for core debris to be ejected into the upper containment regions leading to direct containment heating SECY-90-016: - Provide a depressurization system and cavity design features to contain ejected core debris ACRS LETTER:

- High pressure core melt ejection is an extremely improbable event

- No need to require two modes of coping with it

- Reliable depressurization is probably the preferred approach STAFF RESPONSE: - High pressure melt scenarios are dominant for some 1

designs

- Provide a design with a degree of prevention and mitigation to address uncertainty l

COMMISSION SRM:

Approved the staff's position

- The cavity design should not unduly interfere with operations, l

including refueling, maintenance, or surveillance activities DRAFT CP: - Provide a reliable depressurization system

- Provide cavity design features to decrease the amount of ejected core debris that reaches the upper containment l

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- Lower drywell design to decrease the amount of core debris that reaches the upper drywell

- Uncertainty and sensitivity analysis performed STAFF EVALUATION: - Review of ADS indicates a reliable system

- Containment design limits entrainment into the upper drywell

- Impaction of debris on structures

- Partitioning of gas flow between the upper drywell and suppression pool

- Solid reactor vessel skirt

- Inert containment precludes combustion of hydrogen generation produced from adding to containment pressurization

- Results of review of uncertainty analysis will be provided in PRA evaluation STATUS: This issue is closed i

L 23

I CONTAINMENT PERFORMANCE GOAL PURPOSE:

- To provide a balance between accident prevention and consequence mitigation

- To provide a final check on containment performance through a defense in depth philosophy SECY-90-016: - Use of a conditional containment failure probability of 0.1 or a deterministic containment performance goal that offers comparable protection in the evaluation of evolutionary LWRs CCFP: - Defined as the probability of failure of the mitigation systems (systems which can reduce the consequences of a core damage accident) from the onset of core damage to loss of containment integrity resulting in an uncontrollable leakage substantially greater than the design basis leakage DETERMINISTIC:

- Proposed in recognition of the large uncertainties (including assessment of common-cause failures, human errors, and completeness of analyses and uncertainties in phenomenological behavior) inherent in the quantification of very low frequency l

scenarios being analyzed in a PRA l

- Deterministic containment performance goal The containment should maintain its role as a reliable leak tight barrier by ensuring that containment stresses do not exceed ASME service level C limits for a minimum period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the onset of core damage and that following this 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period the containment should continue to provide a barrier against the uncontrolled release of fission i

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ACRS LETTER:

- ACRS previously recommended such a quantitative guideline for containment performance as part of implementation of the Safety Goal Policy

- Should be used as guidance to the NRC staff in development of requirements for applicants and not passed on to applicants because the definition of CCFP is to imprecise

- Deterministic performance criterion is difficult to interpret

- ACRS has study underway on containment design criteria for future plants but -it was not expected to directly affect the certification of evolutionary LWR designs STAFF RESPONSE: - Staff believes that either a CCFP or deterministic performance goal offers comparable protection' and is necessary to give designers guidance on what is acceptable for containment performance under severe accident conditions COMMISSION SRM: - Approved the use of a 0.1 CCFP as a basis for establishing regulatory guidance

- CCFP should not be imposed as a requirement in and of itself

- CCFP should not discourage accident prevention

- Staff should review and submit to the Commission any suitable, alternative, deterministically-established containment performance objectives providing comparable mitigation capability submitted by applicants DRAFT CP: - Position unchanged GE PROPOSAL: - CCFP of 0.1 for internal events STAFF EVALUATION: - Evaluation of CCFP will be provided in the PRA analysis

-GE has also demonstrated capability of the containment through deterministic evaluation 28 4

ASME SERVICE LEVEL C Service Level C is determined through an evaluation of membrane stresses and buckling of the steel torispherical upper drywell head GE PROPOSAL: - Primary membrane stress controls the design at the Level C Service Limit

- Results of the NASTRAN computer code for calculation of membrane stresses 106 psig at 340 F 97 psig at 500 F

- Buckling failure based on test data Best estimate internal pressure 252 psig Lower bound value 166 psig l

STAFF EVALUATION:

- Using the ALGOR computer code, the staff calculated similar membrane stresses

- Using ASME code provisions to include a factor of safety, the allowable internal pressure for buckling stress is 101 psig or 99 psig

- Pressure capability is controlled by the membrane stresses

- Service Level C limit is 97 psig l

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CONTAINMENT ULTIMATE PRESSURE CAPABILITY GE PROPOSAL:

- Ultimate pressure capability of the containment is limited by the drywell head whose failure mode is plastic yield

- Pressure capability 134 psig at 500 F 120 psig at 700 F STAFF EVALUATION:

- Using a lower bound pressure from tests resulting in failure and a knockdown. factor, the' staff calculated a

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- Staff concludes that the value calculated by GE of 134 psig can be used as the median fragility valug for the drywell head l

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CORE-CONCRETE INTERACTION ANALYSIS PURPOSE:

- Limit the challenges to containment integrity from core-concrete interactions including pressurization from non-condensible gases, generation of combustible gases, liner melt-through, high temperature gases, and structural failure mechanisms SECY-90-016: - Provide sufficient reactor cavity floor space to enhance i

debris spreading

- Provide for quenching debris in the reactor cavity ACRS LETTER: - Agree with the staff's recommendation

- Resolution of this issue will require engineering judgement

- Quantification of sufficient reactor cavity floor space is still an open issue, as is the means by which one quenches the core debris STAFF RESPONSE:

- Evaluation of this issue continues as more experimental data and information become available i

COMMISSION SRM: - Approved the staff's position DRAFT CP: - Provide cavity floor space to enhance debris spreading

- Provide a means to flood the reactor cavity to assist in the cooling process

- Protect the containment liner and other structural members with concrete, if necessary

- Ensure that the containment can accoinmodate a range of core-concrete interaction scenarios for approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> l

31

4 GE PROPOSAL: - !.arge lower drywell floor area (88.25 m ) designed 2

2 greater than.02 m /MWt

- Firewater system provides water to reactor vessel, drywell sprays, and wetwell sprays through RHR system

- Lower drywell passive flooder provides water to lower drywell (melting point 533 K)

- Use of basaltic concrete (1.6 m) for protection of liner 1

- Utilized MAAP for analysis of accident sequences

- MAAP predicts minimal, if any, CCI

- Time to COPS actuation typically ranges from 19 to 30 hrs.

- To address uncertainty in potential for CCl, GE modified MAAP to evaluate continued CCI

- Uncertainty analysis varied amount of core mass, upward heat flux, and other sensitive parameters

- GE analysis indicates COPS actuation in approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />

- Example: Case FMX1 P 100% core mass,100 kW/m upward 2

heat flux, axial ablation (50 hrs) 1.55 m, radial ablation (50 hrs).31 m, COPS actuation 17.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 32 a

6 STAFF EVALUATION: - Cavity sized as large as practical

-Two diverse means of providing water to the cavity

- Concrete provides sacrificial barrier for the liner

- GE has made-a good faith effort-to provide a lower drywell that would promote coolability-

- Design criteria of SECY-90-016 is met

- GE analysis varied key parameters and addressed uncertainty

- GE met criteria to perform aHrange of -. scenarios, but other potentially more limiting cases cannot be ruled out

- Staff independent assessment performed with CORCON-MOD 3

- Preliminary results of GE case FMX1P indicates 1.65 m of axial ablation and.67 m radial. ablation'in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (100 % core mass 2

and 100 kW/m upward heat flux)

- Staff will continue. to investigate CCI to.obtain a better understanding of the containment capability

- At this. time, the staff does not anticipate the need for further analyses from GE

- Staff anticipates final closure of CCI by June 1993 i

STATUS: - The staff is continuing to evaluate this issue b

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l PURPOSE: To ensure that core debris within the lower drywell will not collect within sumps and cause increased ablation CONCERN: Sumps imbedded within the lower drywell could lead to the accumulation of core debris and increased localized ablation rates i

GE PROPOSAL: - Prevent core debris from entering the floor drain sump or equipment drain sump

- GE provided a conceptual design of a corium shield which could be erected around the drain sumps to prevent the ingression of core debris i

STAFF EVALUATION: - Conceptual design should ensure against debris ingression however details have not been provided

- Method of ensuring implementation of conceptual design is under discussion STATUS: - This issue is open 3B

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