ML20206B244

Summary of ACRS Advanced Reactors Design Subcommittee 880803 Meeting in Washington,Dc Re Review of NRC Draft SER for Modular High Temp gas-cooled Reactor & Remaining safety- Related Issues from 880622 Meeting

ML20206B244
Person / Time
Issue date:	08/15/1988
From:	Advisory Committee on Reactor Safeguards
To:	Advisory Committee on Reactor Safeguards
References
ACRS-2594, NUDOCS 8811150386
Download: ML20206B244 (18)

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CERT 1FH DATE ISSUED: 8/15/88 ACRS Meeting Minutes /Sumary of the Advanced Reactors Designs Subcomittee August 3, 1988 Washington, D.C.

Purpose The purpose of the meeting was to continue the review of the NRC staff's draft safety evaluation report (SER) for the modular high temperature <

gas-cooled reactor (MHTGR), and to discuss any iemaining safety-related issues as a result of the June 22, 1988 subcomittee meeting on the same subject.

s Attendees ACRS fGC E'"Va rd , Chairman E Williams, RES W. Kerr, Member J. Wilson, RES F. Remick, Member P. Shewon, Member C. Siess, Member -

C. Wylie, Member J. Lee, Censultant R. Aicry, Consultant D. Okrent, Consultant M. El-Zeftawy, Staff S. Long, Fellow Others D. Moses, ORNL A. Baxter, GA A. Neylan, GA F. Silady, GA C. Rodriguez, GA S. Inamati, GA J. Kendall, GCRA A. Millunzi, DOE R. Mills, PDC 0 S. Poltark, SERCH C. Lewe, NUS L. Walker, Stone & Webster b A RM"R8Rosoocolo 2594 PNV f

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, Meeting Highlights Agreements, and Requests  ;

1. Mr. Ward, Subcomittee Chairman, stated the purpose of the Subcom- .

I mittee meeting and introduced the other present ACRS members and I consultants. Mr. Ward comented tnat the draft SER was incomplete l for the June 22, 1988 subcomittee meeting, however, the staff l currently seems to have completed the missing sections except for  !

the "executive sumary." Mr. Ward indicated that following the f I

subcomittee's review and discussion with the full Comittee, the

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ACRS is expected to write a letter to the Comissioners addressing i this issue. I

2. Mr. J. Wilson, Section Leader / Advanced Reactors and Generic Issues l Branch (ARGIB)/RES, sumarized the SER status as follows:  !

• ARGIB review of the MHTGR is complete.

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' The missing sections of the SER, namely; I Chapter 3 - criteria, l Chapter 11 - radionuclide control, l Chapter 12 - occupational radiation protection, and l r i Chapter 14 - plant and perfonnance testing j were completed and submitted. -

• Completion of internal reviews by RES and NRR are expected by l August 1988. The CRGR is expected to complete its review by September 1988 and then it will be sent to the Comission j approximately by late September 1988.

f Mr. Wilson indicated that there are two Comission papers: i) l Licensing issues (SECY 88-203) addressing accident selection, !t siting source term selection and use, adequacy of containment, and adequacy of emergency planning, and ii) Standardization issues j (SECY 88-202) addressing scope of design to be certified, level of l i

Advanced Reactor Designs Minutes August 3, 1988 9

design detail to be certified, plant options to be certified, and prototype testing, that will be submitted and discussed with the Commission shortly. SECY 88-203 is expected to be discussed on August 9, 1988 and SECY 88-202 is expected to be discussed on August 11, 1988.

3. Mr. P. Williams ARGIB/RES, stated that the Commission expects the safety enhancer. ant that was outlined in the advanced reactor policy statement be achieved. However, this is only an expectation, but not a requirement. Mr. Williams outlined four indicators of safety enhancement as follows:

a. DOE response to staff's questions, concluding:

• slow thermal transiants

• insensitivity to operator errors

• fission produce recention in fuel ender extremely adverse postulated events

• highly reliable passive and inherent safety features for decay heat removal and reactor shutdown

' helium provides a single phase, non-chemically reactir.g coolant

b. Conformance with Advanced Reactor attributes. The staff believes the MHTGR conforms with nine attributes that repre-sent measures of enhanced safety. The staff relics on antic-ipated favorable resolution of certain safety issues and the results of research and development to verify the MHTGR

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Advanced Reactor Designs Minutes August 3, 1988 conformance with the advanced reactor attributes specified n i the policy statement.  ;

) .c. Research program conclusions pertaining to severe accident analysis. The staff indicated that necessary research will be f

identified in critical areas (e.g., fuel integrity, fission

product transpc.. vessel service). The source tenn is [

mechanistic.  !

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!, d. PRA and staff review - PRA results are non-conclusive because l of lack of empirical data base, vulnerabilities from large }

seismic events, and hidden failure mechanisms in unique safety f

components.  !

. Mr. Williams discussed the containment is8ue for tne MTGR.  !

j Presently the MHTGR has no conventiorral containment design and consists of the following:

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* "Tortuous" path for fission product held y and deposition.  :

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• Silo desig:. provides for missile protection that makes sabo-tage more difficult. [

• Remedial action to seal off cavity would appear feasible.

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1 Mr. Williams stated that the NRC itself will decide whether to l i

require the reactor building or an effective alternative to provide l a containment building comparable in structural design and purpose [

I to those of 1.WRs. However, the staff presently agrees with the {

DOE's position that, in principle, the reactor building may not l need to provide pressure retaining capability if the design objec- l tives for the fuel are met.

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Advanced Reactor Designs Minutes August 3, 1988 Mr. Williams commented that the staff's report entitled "Risk Benefits of a Containment System for the MHTGR" (dated May 1988) illustrates design options for a pressure-retaining containment oesign that would offer a more conventional function than that proposed by DOE. This report evaluated three alternate containment options as follows:

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• A water-cooled pCCS with a containment tunnel for venting the l reactor silo. This design is similar to a design proposed for I DOE's New Production Reactor to produce tritium.

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• A contain1ent building enclosing the entire air-cooled RCCS l

with an additional cooling system for the containment atmo-sphere.

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• An air cooled RCCS penetratir.g a containment building with adequate provision for isolating the RCCS.

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( The repnrt indicates that the first two options decrease the capability of the plant to prevent a large radionuclide release because of the decreased reliability of the decay heat renoval I systems. The report concludes, if a traditional containment function is desired, it should be accomplished with the third option.

Mr. Williams presented the prototype plant testing issues. For the MHTGR, DOE proposes a standard plant demonstration at a utility site. This it in contrast to the staff's position that testing and demonstration of a reactor plant without conventional containment ,

must be performed at an appropriate isolated site.

The 00E proposed demonstration plant is primarily for commercial demonstration with initial startup data to be used to confirm l

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specific aspects of design and does not provide for specific safety tests.

The staff envisions the prototvse reactor as a single module, with

' associated instrumentatic.i and controls and other systems important to sr.faty at least through and including the steam generator. The purpose of the' test program would be to verify the analysis of the plant response to bounding events and generate experimental data to verify the analytical tools. The staff indicated that testing that damages the plant would not be necessary.

Mr. Williams outlined some future plans that includes the follow, ing:

Evaluation of progress in DOE research and development pro-g rams ,

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• Evaluation of prototype plant testing plans.

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• Review available information on structural graphite and core 4

support structure. <

• Evaluation on additional topics identified in SER (e.g.,

frequency of elevated vessel temperatures, vessel system seismic design, pressure relief system, etc.).

i Based upon the results of the review, the staff concludes that the MHTGR design has the potential to achieve a level of safety equiva-lent to current generation LWRs and, in some areas, provide safety enhancement beyond that in current generation LWR designs. The main factors leading to this conclusion are the following:

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Advanced Reactor Designs Minutes August 3, 1988

• The potential for only minor core damage and fission product release over a wide range of severe challenges to the plant.  !

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• The reduced dependence on human actions and the reduced [

vulnerability to human error, i f

• The long response time of the reactor under accident condi-  ;

tions, providing time for evaluation end corrective action. [

• The capability to demonstrate by test the significant safety features and perfonnance of the plant over a wide range of j events. t 1

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• The results of indepandent analysis by ORNL and ML which [

i indicated good agreement with the designers predicted perfor- r 4 -;

reance.  ;

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f 4 Mr. O. Moses, ORNL, sumarized the independent analyses performed f I at ORNL. He stated that from the LOFC heatup occident analyses, (

l the MHTGR design is not suscepP,ible to significant fuel failure s c

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n costulated accidents. One major area of concern is with j posrible vessel overheating. Sensitivity studies showed that the l most crucial safety-related parameters or operational uncertainties j j were the core thermal conductivities, the afterheat curve, and the  !

effective RCCS heat removal performance. Mr. Moses indicated that i i for the reactivity insertion studies, it could be concluded that f given the current nuclear parameter functions as input, the results l j of the postulated accidents are acceptable. Independent checks should be made how ver, on the reactivity worth of steam in the i

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Advanced Reactor Designs Minutes August 3, 1988 4

core, the effect of core moisture on control rod worth, and mechanisms for more massive moisture ingress rates.

Mr. Moses pointed out the major licensing short fa'11s as follows; N

• The general lack of good data from neutronic experiments reletant to the MHTGR core design,

• The lack of documentation and quality-assured analyses.

• The lack of independent analyses by disinterested parties.

Mr. Moses commented that the' situation faced by the NRC is as I follows:

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• The development of independent confirmatory analytical ca-l pability is expensive (perhaps as much as $1M to $2M),

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• DOE has promised a nuclear design section in the RTDP and overseas cooperative experiments.

- NRC can insist on continuing DOE /NRC contact.

- NRC can visit / audit overseas facilities (Regulatory Guide I 1.68; ANSI /ANS-19-series standards).

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• NRC can probably expect DOE to put off design assurance to first unit startup testing.

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' - This will not address water ingress effects.

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- E0EC temperatura coefficient negativity will not be ccnfirred until after several years of operatioa.

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Advanced Reactor Designs Minutes August 3, 1988

5. Mr. A. Baxter, GA-Technologies, sumarized the neutron control strategy in the design as follows: l i

• Rods are moved in groups of 3 rods (at 120' synnetry) for  !

normal control including rise-to-power, load-following and -

burnup compensation, i

• The two groups (6 rods) in inner reflector are wtthdrawn first l

! in rise-to-power - fully withdrawn 9 9 25% nominal value, f

• The 8 groups (24 rods total) in outer reflector are sequen- t tially withdrawn to reach full power conditions.

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• Normal full-power conditions require one to two groups of rods l

to be inserted typically to control 4 3% sn hot excess. '

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• liner rods prohit.ited from scram at > "25% full-power after j one outer rod group has become fully withdrawn, t r

• Hot shutdown achieved by trip of all outer reflector control i rods (as a bank of 24 rods), t I f

! ' Cold shutdown achieved by trip of outer rods 'p/or [

] insertion of inner 6 rods.  !

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• Reserve shutdown control (RSC) inserts boronated pellets into f 12 active core locations. Automatically inserts on de'1yed l power / flow trip or on overpressure trip after outer ros trip (

for moisture ingress. May also be inserted manually, j i

l Mr. Baxter indicated that, there is an adequate control and shut- l down nargin under all conditions. I I

I Advanced Reactor Designs Minutes August 3, 1988 The sumary of the MHTGR reactivity feedback components is as follows:

• Large, prompt, negative doppler feedback.

• Large, negative moderator temperature feedback.

• Negligible feedback from coolant density or pressure changes (total lots of helium is worth less than le in reactivity).

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• Negligible feedback from graphite dimensional changes.

• Negligible feedba;k from vessel dimensional changes, t

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• Small, delayed, positive feedback from reflector temperature cl.anges through neutron leakage effects.

l Mr. Baxter indicted the following for the MHTGR design:

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• Power coefficient always negative.

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' Temperature coefficient increasingly negative above and below l operating temperatures.

• Stable reactor characteristics.

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• Behavior similar to and verifi'ed by tests on operating AGRS

! HTGRs(AVR,PB,FSV).

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6. Mr. F. Silady, GA Technologies, sumarized the role of operator i requirements as follows:

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• No operator actions required to meet safety requirements.

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• Plant safety insensitive to operator errors. f f

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• Concentrate efforts of operator on efficient and economic }

plant operation.  !

3 Mr. Silady indicated that due to the MHTGR inherent characteristics  ;

j and passive safety features, the need for operator safety actions l' l

is eliminated. The staff accepts, with caution, the proposed fully

, automated control system and the control of the four reactor  !

l modules by a single plant operating crew. However, the staff does  :

! not agree with DOE's pioposal that she role of the operator is not  !

l safety related.

r l 7. Mr. F. Silady, GA Technologies, described the importance of  !

graphite oxidation for the MHTGR. He stated that the NRC consul-  !

l tant has confirmed DOE calculations that the MHTGR core design

{ precludes significant graphita oxidation leading to a self- 1 I

sustainad reaction or unacceptable temperatures. GA has examined a i

! broad spectrum of air ingress scenarios with the conclusion that I graphite oxidation has negligible 4mpact on MHTGR's acceptably low

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j consequences.

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8. Mr. A. hylan, GA Technologies, presented an overview of the l technical issues regarding the MHTGR design. He stated that DOE

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j and the NRC are generally in agreement that technical issues  ;

{ identified in the SER can be adequately addressed in subsequent  !

design and technology development programs. Specific technical f areas for continued DOE /NRC review are: ['

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• Accident selection: continue development of systematic l

' approach to event specification and use, f-1 k

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Advanced Reactor Designs Minutes ' August 3, 1988 I

Regulatory criteria: continue development to top down compre- l hensive rationale for licensing bases. l Uncertainties: assure design and technology is fo(;used on key  !

i teres. ,

Vessel temperature > 700*F: available data indicates accept-f 4

able performance. Code inquiry initiated.

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Vessel reliability: investigate potential failure modes to i assure reliability 2s LWR vessel, f Safety classification: component design bases, QA, reliabil- l ity to be defined.

Fuel testing and QA: agree to further define scope / ,

i requirements to expand statistical basis for fuel failure /

) fission product transport models and fresh fuel QA.

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l Reactor physics: agree to document / validate computer codes; f i quantify design margin; revise RTDP as required.

i l i Mr. Neylan emphasized that the MHTGR safety design approach assures l

the radionuclide retention within fuel with simple passive design l I

features, and the licensing basis events are syttematically select- I

! ed. He conrented that the MHTGR position on adequacy of contain-  ;

j ment system is consistent with rational licensing approach and l intrinsic, physical characteristics of MHTGR. He also indicated

that the MHTGR does not require prototype safety tests for licens-j ing. I I

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Advanced Reactor Designs Minutes August 3, 1988

9. Mr. A. Neylan, GA Technologies, answered the questions concerning nuclear grade graphite properties that was provided by Jack Parry as follows:

t Q.1 Is the base coke a calcined petroleum coke?

A Yes.

Q.2 Can a reliable source of the same coke be assured for succes-sive reactor modules? If so, how? (Note: Coke varies with oil field, refining method; season of the year, and weather if stored unprotected in bulk piles.)

. A. Yes. More discussion with graphite vendors is expected, i

Q.3 How is the graphite formed (extrusion, pressing, or isostatic forming)?

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A By extrusion.

Q.4 For a full size billet, what are the following graphite i properties:

(a) Coefficient of thermal expansion (longitudinal and i

transverse)?

(b) Electrical resistance?  :

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(c) Compressive strength?

(d) Pore size distribution? ,

(e) Degree of graphitization? [

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Advanced Reactor Designs Minutes August 3, 1988 h It is measured by all the above properties.

,3 How 10 the bove properties vary within a billet? From billet L4 tie *0 h Within a billet, it is stronger on the outside and weaker la

. From billet to billet is strength and densi-

ty.

Q.6 What 1rradiation studies have been done on the proposed grade of graphite? For completed studies, were te3t specimens from a full size billet or from lab prepared specimens?

A2 It is done on statistical basis and Q/A plan, and the samples were taken from a full si.:e billets.

Q.7 Which aspects of core design for the MHTGR are most sensitive to variations in graphite properties? What safety issues, if any, are related to these P0tential property variations of graphite within a single core or between different modules?

h The most sensitive properties are strength (variation from possible cracking) and conductivity (decreases with radiation),

10. Mr. A. Millunzi 00E, following the conchsion of the Subcomittee meeting, urged the Subcomittee members to approve the conceptual design or the MHTGR based on the technical aspects rather than the political ones and waive the policy matters to a later date.

11. As a result of the Subcomittee discussion, some of the Subcommit-tee's members and consultants expressed some concerns in regard to the following:

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' Dr. Kerr questioned the meaning of tha statement made by the NRC staff that indicates the Commission's expectation of safety enhancement for the MHTGR design, but it is not required.

• Dr. Shewmon expressed some concern it. regard to the seismic design of the core and the need to verify the graphite prop-erties (e.g., toughness) to ensure consistency and structural stability.

• Dr. Shewmon expressed some concern in regsro to the verifica-tion of vessel failure mechanisms. In addition, the assump-tion of low probability risk of vessel failure has to be justified under conditions of pneumatic overpressure or overtemperature,

• Mr. Ward c.uestiened the role of reactor operator for this design. He pointed out that the two major power reactor acciderts that have occurred, TMI-2 and Chernobyl were caused by wrong but purposeful actions by operators. PRA does not do a good job in describing or estimating the likelihood of such events. If the MHTGR is claimed to be safer than LWRs, then there should be some systematic means to analyze its resis-tance to such maloperation.

• Or. Kerr questioned the decisions and assumptions that were made by the NRC staff to account for the defense-in-depth philosophy.

• Dr. Remick commented that there is no accident scanario in the analysis that includes flooding the reactor cavity and the recovery process.

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• Dr. Remick questioned the various means by which the NPC Staff reviewed the neutron control and shutdown margin under various  ;

conditions.

• Dr. Xerr commented that it is not clear that if the conceptual design can deal with the station blackout issue for the longer time duration specified by the Staff, and additional analysis should be performed to validate that assumption.

• Dr. Kerr questioned that if the external events were included in the IRA studies in a proper manner.

• Dr. Terr expressed some concern in regard to the flow blockage and the lack of analysis and experimental data to support the staff's argument that this event is considered to be remote.

However, the staff responded by stating that, if blockage became extensive, significaitt fuel failure could be detected by gaseous fission product activity monitors in the primary system.

• Dr. Sicss questioned the fuel particle failure rates as-sumptions that were made by the Staff, and added that there is no experimental available data to validate those assumptions. l

• Dr. Remick concented that the containment issue need q be discuss *d further with emphasis on the public percention.

• Dr. Okrent expressed strong concerns regarding the following:

(a) The NRC's staff interpretation of defense-in-depth is f confusing.

i (b) The PRA study is incomplete and highly subjective.

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4 (c) Flood vulnerability should be examined. >

j (d) The seismic analysis is subjective and incomplete. l, I

(e) More proof of inspection capability and reliability is ,

needed. i (f) The frequency of plane crash may be high enough and ,

l should be examined. f

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(g) Yessel failure effects should be expanded and examined, f i

l (h) Gross steam generator failure should be examined, f i  !,

(1) Cost / benefit analysis could be inadequate to justify the l alternate containment options. l l  !

(j) PRA uncertainties are inadequate.  !

(k) The analysis of structural failure possibilities is l inadequate.  !

1 (1) Safeguards vulnerability should be examined. f l

• Dr. Lee commented that the MHTGR reactor physics methodology [

t appears to be in need of substantial verification and possibly l

further developments. In particular, the proposed low enrich- l

/ ment uranium / thorium fuel configuration may represent an l intermediate neutron flux spectrum, for which only limited i experience in reactor physics exists. Dr. Lee will send additional comments at a later date.  !

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• Dr. Avery **.perssed a great deal of concern in regard to the containment issue. He commented that the NRC staff and DOE rely heavily on the argument that it is not possible for significant activity to escape the fuel particles. Dr. Avery added in his views, this has not been shown and proven yet.

Dr. Avery will supply his additional comments later on.

Future Activities The Subcommittee Chairman is planning to brief the full Committee, in

August 1988, regarding the Subcommittee activities. In cddition, the j NRC staff and DOE representatives will brief the ?.ommittee regarding the NHTGR conceptual design. The ACRS may wish to write a letter on this subject.

NOTE: Additional meeting details can be obtained from a transcript of this meeting available in the NRC Public Document Room, 1717 H Street, N.W. , 'e shington, D.C., or can be purchased from Heritage Reporting Corporation,1220 L Strest, N.W.,

Suite 600, Washington, D.C. 20005 (202) 628-4888.

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