ML20206A337

Summary of ACRS Advanced Reactor Designs Subcommittee 880106 Meeting in Washington,Dc Re Review of Draft Commission Paper on Severe Accidents & Containment Issues for DOE- Sponsored Advanced Reactor Designs

ML20206A337
Person / Time
Issue date:	01/14/1988
From:	Advisory Committee on Reactor Safeguards
To:	Advisory Committee on Reactor Safeguards
References
ACRS-2545, NUDOCS 8811150070
Download: ML20206A337 (11)

Text

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OARS-S HE

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I'g \ !s DATE ISSUtD: 1/14/88

.; ter MFETING MINUTES /

SUMMARY

FOR THE ADVANCED REACTOR DESIGNS SUBCOMMITTEE MEETING JANUARY 6, 1988 WASHINGTON, D.C.

I POSE 1

The purpose of the meeting is to review and comment on the draft Commis-sion paper regarding the severe accidents and containment issues for the

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DOE-sponsored advanced reactor designs. l i

ATTENDEES ACRS NRC Staff, c 1

D. Ward, Chairran T. King, RES l J. Ebersole, Member Z. Rosztoczy, PES  ;

J. O. Mark, Member J. Wilson, RES C. Michelson, Member P. William, RES ,

F. Remick, Membar J. Flack, RES i

.l C. P. Siess, Mettber R. Landry, RES i J C. Wylies Member B. Hardin, RES R. Avery, Consultant M. Spangler, NRR  ;

P. Davis, Consultant M. Dey, RES M. El-Zeftawy. Staff G. Brown, Fellem A. Tabatabai, Fellow l

5. Arndt, Fellow '

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Others ,

I G. Gyorey, GE R. Lancet, RI J. Marchatene, ANL F. Silady, GA C. Lewe, NUS F. Stetson, NUS

W. Bruss, Bechtel J. Maffre, USCEA

P. Beliveau, NUMARC G. Sherwood, DOE J. Scarborough, OCM N. Klug. DOE ,

O. Pedersen, ANL F. Gavigan, DOE  !

W. Pasedag, INEL i l

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. 8811150070 880114  ;

PDR ACHS t 2545 PNV j DMK  :

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Advanced Rea tor Designs Minutes January 6, 1988 MEETING HIGH'IGHTS, AGREEPENTS, AND REQUESTS

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1. Mr. Ward, Subcomittee Chainnan, stated the purpose of the Subcom-mittee neeting and introduced the other present ACRS members and  ;

consultants. Mr. Ward comented that due to the unique charac-teristics of the DOE-sponsored advanced reactor concepts, the approach to key licensing issues does not exactly follow that of conventional LWRs.

2. Mr. Z. Rosztoczy, Deputy Director / Division of Regulatory Applica-tions/?ES, stated that the NRC Staff developed an approach and  !

criteria which utilizes the guidance in the Commission's advanced  ;

reactor, sa'ety goal and severe accident policy statements as the i basis for a decision making. These criteria are generic in nature, and will be t rovided for Comission review via a SECY paper (draf t i provided to fiubcomittee). Mr., Ros2toczy comented that prior to i sending the SECY paper to the Comission, the ACRS and CRGR reviews are requested. There are two draft Comission papers that were prepared by the NRC Staff; one regarding the key issues, and the

ot.her regarding the standardization. The Staff is planning to send both papers to the Comission by late March 1988.

i 3. Mr. T. Kirig. Chief / Advanced Reactors and Generic Issues Branch /RES, j presented a summary of the DOE-sponsored advanced designs. These 3

are: (1) 350 Mwt mod slar HTGR, (2) 425 MWt modular LMR (PRISM),

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and (3) 900 MWt redult r LMR (SAFR) designs. For approximately the

) past year, the NRC Staff has been reviewing those designs.

The HTGR Concept:

• Designers - GA Technologies /S&W/Bechtel/CE/GE.

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• 350 Mwt (137 Mwe), modular design, with reactor vessel and l

steam generator located below grade.

s Advanced Reactor Desig~ iiinutes January 6, 1583 :s

• Steel reactor vessel.

• External steam generator.

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• TRISO coated fuel particles similar to FSV. <

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• Annular core design with prismatic fuel blocks similar to FSV.

• One loop per module.

' one turbine-generator per 2 modules.

• No containment or confinement building.

• Passive decay heat removal and shutdown systems.

• None-safety grade BOP.

• 40 year module lifetime. -

The PRISM Concept:

' Designers - GE/Bechtel/ United Engineers / Byron-Jackson / Foster-Wheeler.

• 425 Mwt (138 Mwe), modular design, pool type LMR with reactor vessel and stehm generator located below grade.

• Heterogeneous core design with metal fuel.

' One loop per module.

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• One turbine-generator PEP 3 modules, f;

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• Reactor guard vessel used as containment boundary. [

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• Fassive decay heat removal and shutdown systems.

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• Non-safety grade BOP. li 4

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• 60 Year trcdule life.

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• Utilizes idea of safety demonstration test to facilitate (

1. licensing / private sector acceptance.  !

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The SAFR Concept

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' Designer - Gockwell International /Bechtel/CE. f

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• 900 Mut (350 Mme), P0dular design, p001 type LMR with reactor {

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vessel and steam generators above grace.

' Heterogeneous core design with metal fuel, i

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' Two loops per module.

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! ' One turbine-generator per module.

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• Reactor guard vessel used as contair.nent beundary. [

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• Passive decay heat rsmoval and shutdown systems.

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• Non-safety grade BOP.  ;

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Advenced Reactor Designs Minutes "t January 6, 1988

• 60 year module lifetime. 'i 8 rd a

Mr. King mentioned that the St'ety Evaluation Report (SER) regard-ing the MHTGR is alaiost complete (draft) and for the LMRs. the drafts would be completed by late March 1988. Mr. King indicated also, that the Staff is planning to have ACRS review the SERs prior to trar.

mittal to Comission.

In regard to the Commission Paper on key licensing issues, the

overall goal tf the NRC Staff approach is to ensure, consistent with the guidance of the policy statement, that advanced reactors achieve a level of safety at least equivalent to that of current

] generation LWRs. The proposed gentral criteria for advanced reactors is as follows:

) - Feet all applicable regulations (10 CFR/SRP).

- Comply with severe accident policy statement.

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- Comply with the Commission's safety goals (including the guidelines that the probability of a large release should be less than 10 I/ year).

- Provide two redundant, diverse and independent means of reactor shutdown and decay heat removal.

- Verify plant perfonnance via prototype or first plant testing.

- Verify via cost / benefit analysis, that alternative design approaches in key areas are not justified.

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, Advanced Reactor Designs Minutes C .ary 6, 1988 t

The draf t Comission Paper presents a set of cri which the NRC Staff proposes be used to assess the DOE-sponso: Ivanced reactor l concepts in the areas of-(1) Accident selection. [

(2) Selection of a site suitability source ters. f (3) Adequacy of containment. l

,(4) Adequacy of offsite emergency planning. i s

Due consideration of uncertainties was factored in. The specific criteria for the above four key areas is as follows:  ;

(1) Select a set of design basis and beyond design basis events  !

for each plant utilizing engineering judgment supplemented by [

PRA. Select accidents conservatively to bound uncertainties. l Look at accident sequence down to 10~7/ yea r. '

(2) Select site suitability source tem (SSST) as source from worst accidents.

(3) Evaluate adequacy of containment based on SSST above and must wet 10 CFR 100 dose guidelines.

(4) No preplanned offsite evacuation would be required if radioac-tive releases from the plant do not eneed the EPA PAGs during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of an accident.

The NRC Staff recowerds that the designers for the advanced reactor concepts be required to assess their designs for compliance with the above criteria at each stage.

Advanced Reactor Designs Minutes January 6, !!88 l

Mr. King described the NRC Staff approach for defense-in-de t requirements. The requirements address the different stages and aspects of plant safety which can be generally categorized as:

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• Prevention

• Protectiot.

• Mitigation

• Emergency Planning Mr. King stated that the advanced reactor designers have approached plant designs and the means of maintaining defense-in-depth some-what different than the approach taken by LWR designers. In general, the advanced reactor designers make a shift in emphasis from mitigation features to highly reliable protection features.

l 4 Mr. R. Lancet, Rockwell's International (RI), stated that RI's l approach is to license the first plant (LMR - SAFR design) due to f the fact it would be economically viable plant and produces power.

j However, as a part of the first plant, a series of safety tests and

demonstrations would be run first. The design certification will j be obtained after the first plant has been opersted and the safety

! tests were performed. So, there would be a licensed prototype i

before granting the final design certification. The design that would be licensed is a single module (one reactor and one turbine).

For the PRISM reactor, the same approach basically applies except

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1 the first plar.t is not licensed.

5. Mr. G. Gyorey, General Electric Co., (PRISM designers), comented j

that for the PRISM, the proposition is to build just a nuclear l island portion of PRISM on an existing DOE site, then run all the j safety tests to demonstrate conclusively the inherent characteris-I tics of both shutdown and decay heat removal. However, to make

! that plant practical in the long tem, a turbine will be added to I

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. ," ' Advanced Reactor Designs Minutes January 6, 1988 i form a single unit. After the safety tests of the nuclear island ,

portion of the single unit were perforted, then propose to proceed l to design certification for the full plant. (This GE's proposa*v ~

could also run in parallel and may follow somewhat the formal licensing of this initial prototype.)

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6. Mr. F. Gavigan. 00E, comented that DOE and its subcontractors i along with the NRC Staff are focusing on the technical issues of the advanced reactors conceptual designs. He urged the Subcomit-tee members not to get heavily confused with all the definitions (

and get into the argument of deterministic versus uncertainties vs. {

probabilistic vs. safety goals and engineering judgements, and so  !

forth. He commented that the NRC and 00E have a good opportunity to develop and review new reactor designs that are different, safer j and with larger margins, even though the Europeans do not follow )

the same route.

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7. As a result of the Subcomittee's discussion. the Subcommittee  ;

members raised some concerns regarding the following: [

• Mr. Michelson expressed some concern regarding the lack of L knowledge and experience and data base that exist to support i the PRA approach as the principal method for making decisions l for the advanced designs. The NRC Staff agreed. [

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• Mr. Ward questioned the Staff's approach to apply the Com- (

mission's Safety goals for early and latent fatalities, and {

large release criteria on a per site basis for advanced  ;

reactors and not to a single reactor as is the case for LWRs  !

t and this could become a future limitation for NPP "parks", j Dr. Remick shared the same concern. The Staff replied that this would result in a are conservative goal for each reactor

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, Advanced Reactor Designs Minutes January 6, 1988 (

module for non-comon cause events, by about a factor of 10 depending on the number of reactor modules in a site.  !

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• Dr. Remick expressed some concern regarding the issue of no  ;

conventional containment for advanced conceptual designs and [

! the public acceptance of this criteria. Dr. Remick consented  ;

that this approach will require extensive testing and proto- i

types to provide enough confidence for the public to accept [

this major change in the traditional way of building nuclear reactors.

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• Mr. Ward concented that in the draft Corsission's Paper, there l' 1s a requirement that designers will assess their designs for compliance at each stage. The purpose is to assure that new (

information and understanding of the design is accounted for. l

! However, it is possible that actual criteria might need change [

j for the same reasons, and it might be a good idea for there to i be an explicit milestone to recognize this.

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' Dr. Mark expressed some concern regarding the definition of l

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j severe accidents in the advanced designs and how is the severe accident policy would be applied. 1

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• Mr. Ebersole expressed some concern regarding the lack of .

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consideration of air intrusion and harsh environment into the -

I core region in the Staff's evaluation of accidents. The Staff claimed that these conditions are highly unlikely to exist. i l

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• Mr. Ward coerented that the Staff should have a specific plan l l to confinn the uncertainties in the advanced reactor designs.

Other ACRS members shared the same sentimant.

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, Advanced Reactor Design.; Minutes January 6, 1988 i

• Dr. Mark commented that it is not known if it's acceptable to ,

exclude events with a frequency of less than 104/ year from i consideration in decisions regarding source term and contain- I

. ment. [

• Dr. Mark expressed some concern regarding the sabotage issue f j for advanced reactor designs and the actual resistance of

) these plants to such activities. Mr. Michelson shared the j same concern.

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• Dr. Siess questioned if it is acceptable to select a siting i

! source term based upon mechanistic analysis in lieu of an i approach censistent with 10 CFR 100. The NRC Staff will reply I to this question at a later date. [

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• Mr. Ward questioned the separation between Design Basis j Accident (DBA) and Beyon6 Design Basis Events (BDBEs), and if j the designers will really make that distinction. The Staff (

responded that the 800Es will be used for the purpose of site [,

4 suitability source term selection and containment and emergen- i j cy planning evaluation. The designers have proposed these

) BDBis be selected as follows:

. - MHTGR .

Using PRA results at tse conceptual design stage, acci-  !

I dent sequences that have a probability between 10 / year f

[ and 5x10'I/ year are called PDBEs. Events in this  !

I probability range are selected by evaluating accident f

sequences down to 10'8/ year. Uncertainties in their  !

probability are estimated and if the estimated uncertain- !

l ty causes the event to frequency to exceed 5x10'7/ year then the event is considered a BCBE.

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' Advanced Reactor Designs Minutes January 6, 1988 f

- 1.F.RS ATWS events under transient overpower (rod withdrawal),

loss of forced circulation and loss of normal heat sink conditions are considered the BDBEs. These were the CRSR EDBEs.

• Pr. Ward expressed some concern in regard to the Staff's choice for the DBAs which does not include sodiun or graphite fires. The Staff replied that because it mainly relies on probabilistic argument.

FUTURE ACTIVITIES The Subcomittee Chainnan will brief the full Comittee on February 1988 regarding the Subcomittee activities. DOE representatives, its subcon-tractors, and the NRC Staff will all be prepared to give presentations on the description of advanced designs and the Comission Paper at the February 1988 full Comittee meeting. The Comittee may then decide to write a letter regarding the draft Comission Paper on severe accidents and containment issues. Currently, as a result of the Subcomittee meeting, the NRC Staff is incorporating the ACRS coments in its draf t of the Ccomission Paper.

NOTE: Additional rceting details can be obtained from a transcript of this meeting available in the NRC Public Document Room, 1717 H Street, N.W., Washington, D.C., or can be purchased from ACE-Federal Reporters, 444 North Capitol Street, Suite 402, Washington, D.C. 20001.(202)347-3700.