ML20236L276

Discusses 871015 Meeting W/Doe Re Project 673 Concerning Safr Accident Analyses Given in Chapter 15 & App a of Safr Preliminary Safety Info Document (Psid).Meeting Agenda,List of Attendees & Questions Encl.Response Expected by 871130

ML20236L276
Person / Time
Issue date:	11/02/1987
From:	Morris B NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
To:	Gavigan F ENERGY, DEPT. OF
References
PROJECT-673A NUDOCS 8711100247
Download: ML20236L276 (11)

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Hr. Francts X. Gavigan,-Director q

Office of Advanced Reactor Programs  ;

Office of Nuclear Energy U. S. Department of. Energy j Washington, DC -20545

Dear Mr. Gcvigan:

On October 15, 1987, members of the NRC staff and its contractor (BNL) met i with members'of DOE and its contractors to discuss the SAFR' accident analyses  :

given in Chapter 15 and Appendix A of the'SAFR Preliminary Safety Information Document (PSID) - Project 673. The agenda and list of attendees are given in.

Enclosures 1 and 2, respectively. The meeting consisted principally of j presentations by DOE's primary contractor.(Rockwell International .RI) for i this project on the accident analyses described in Chapter 15 and Appendix A of the PSID.

The NRC staff comments on these documents were also discussed. if The action items, including requests for additional information and/or clarifi-cation, that resulted.from the meeting are given in Enclosure 3. A set of preliminary questions were submitted informally. prior to the October 15. 1987 meeting. These provided aLbasis for structuring the meeting and were discussed in some detail at'the meeting. There have been some minor changes to some of these questions to clarify and, in some cases, to expand on the question.

Such questions are marked with an asterisk. Several additional questions resulted from discussions at the meeting. These a e also included in Enclosure 3.

I have also included (in Enclosure 4) questions we recently received from the Radiation Protection Branch of the Office of Nuclear Reactor Regulation. These questions address severel aspects of waste disposal and radiation protection associated with the SAFR design.

I have transmitted to you under a. separate cover a set of additional questions-that we recently received from the Safeguards Bratich of the Office of Nuclear Reactor Regulation. The response.to these questions should be protected-in )

accordance with the requirements of 10CFR50.34(e) and 10CFR2.790(d).

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Enclosures:

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1. Meeting Agenda v.f -(%_,. ~

2. t.ist of Attendees '

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3. Questions f,.i-

4. Waste Pisposal & Radiation ~ Protection n'I'j! ' '

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. Enclosure'1 PROPOSED AGENDA' SAFR PSID CHAPTER 15 AND APPENDIX A' REVIEW ,

OCTOBER.15,-'1987 9:00 a.m. Introduction ~ .R. T. La'ncet. '.

9:15 Design Approach to'. Safety ,J.'C.' Mills 9:40 Reactivity Insertion and Undercooking Events <

10:30- Break- ,  ;

10:45 Primary Pipe Rupture (1 of 8)

Primary Pump. Seizure with Trip Gas Bubbles Through'the Core

-11:25 Accident Without-Scram' Status.

of' Uncertainty Assessment- D.' Wade (ANL) i 12:00 Noon Lunch ' I 1:00 p.m. LOF I

j LOHS l r -

TOP

1 Combined. TOP /LOF Pump Seizure 2:00 Sodium Fires R. T. Lancet '

2:30 -Break 4

2:40 Steam Generator Tube. Break Response H. Chung (ANL)

DBA With Steam / Water Dump DBA W/0 Steam Water Dump BDBA-(Multiple Tube Failures) W/0 Steam / Water Dump 3:20 SSST J. C. Mills 3:50 Sumary R. T. Lancet-4:00 Adjou n v.

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i Enclosure 2

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SAFR Chapter 15 Meeting

-October 15, 1987

. DOE Conference Room -.B005 Germantown, Maryland Name Affiliation- Telephone l

C.. Allen 1 NRC/RES/ARGIB (FTS) 492-8302 1 R. Landry NRC/RES/ARGIB 492-4914' I R. Dale Rogers Rockwell (818) 700-3318 J..N. Wilson RES/NRC/ARGIC (FTS) 492-4727-Dean R. Pedersen ANL-E (312) 972-3335 j

Gregory C. Slovik BNL (515) 282-7983 L ,

(FTS)'666-7883 Greg VanTuyle BNL (516) 282-7960 (FTS) 666-7960' 1 T. King

• NRC/RES/ARGIB (FTS) 492-7014 l J. Flack NRC/RES/ARGIB '443-7767 i R. T. Lancet Rockwell (818) 700-3646 George Sherwood DOE (FTS) 233-4162-Bing C. Chan BNL (516) 282-7166 Jim Cahalan ANL/ RAS (312) 972-4682

, L. N. Rib NRC Consultant' (301) 983-1032 Howard H. Chung ANL/MCT (312)~972-6159 Ronal A. Wigeland ANL/ RAS (312) 972-4683-T. Wei ANL/ RAS .(312) 972-4688 Thomas J. Moran ANL/AP (312) 972-5901  !

S. 4,;1alid Shaukat NRC/RES (301) 492-7821 F. X. Gavigan* DOE

' John Minns NRC/DRPEP (301) 492-4086 i L Al Frost

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Enclosure 3 Request for. Infomation on SAFR Chapter 15 15.1 On page'15.1-4 (Rev. 8),'the phrase."non-safety-grade SASS" appears in the last paragraph.. Clearly the. only purpose of the SASS system.is -

safety. Can:we interpret this as' meaning that.RI anticipates' problems I

in making SASS: safety-grade, and if so why?

l 15.2 There is.a footnote (a) on Table.15.11 3, and a limited qualification?

' statement is' made about .some " isolated components". possibly needing= 'l upgrading. Specifically, which components are marginal,in this regard?

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1I 15.3 In Section 15.2.2.2.3, some pipe break analysis-is described 'However, the wording regarding the flow rates is confusing; First, was~one of theeightpipesrunningfromthe(two)'pumpstothecoreinletjlenum. ]

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ruptured? Second, what fraction of full flow was provided to the.

reactor? Third, how was this fraction determined?-

15.4 In Section 15.2.2.3.2, which discusses inherent response to a postulated pump-seizure event, same "further-SASSYS" analysis is-mentioned (Rev. 7). Has there been any progress on_ this analysis?.

15.5 In Section .15.2.4.4.2, the ' reactivity of gas bubbles in the core is discussed. Analysis from CRBRP is cited in' stating that the' resulting power and temperature increases would be small. .However, because CRBR was to use oxide fuel and thus would have had'a much stronger Doppler feedback, it was less sensitive than a metal fuel-reactor for a reactivity insertion. How large a reactivity insertion was cited in the CRBRP analysis, and how large an insertion would comparable bubbles cause in SAFR7 15.6 A sodium water reaction was analyzed using SWAAM,'as discussed in.

Section 15.2.5.6.2, for one double ended break followed by two resultant -

breaks 0.1 s later. Given the high pressure in the steam system'(2930-psia),' rupture of more than the two subsequent tubes seems quite possible.

How was this number of ruptures chosen?-

I i 15.7 The difficulties cited in Section 15.2.7.1' rerarding identification of-an acceptable SSST event are noteworthy. The staff is in the process of formulating an approach for the selection of this SS5Tl utilizing bounding events.- In this regard, it is requested that a postulated primary pump seizure! combined with'a failure to scram be analyzed and the results provided for information. Was this event considered too  !

unlikely for treatment in the PSID, or were the consequences determined to insignificant?

15.8 The PRA was used to separate DBEs from BDBEs, and thus'it eliminates several (hopefully) improbable events from the Chapter 15 analysis.

Given that there is greater uncertainty regarding reliability factors for new reactor types, how has RI assured that marginally improbable events are considered - just in case the reliability turns out to be overestimated?

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15.9 As eutectic formation is a time dependent process, a brief violation of the threshold temperature is probably acceptable. How is the time-dependent consideration factored into RI's decision making process (i.e.,insettingacceptancecriteria)?

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• 15.10 New results for the unprotected accidents analyzed in Appendix A of the PSID were presented at the October 15, 1987 meeting. In this regard, please provide your most recent estimates for the following metal fuel parameters:

(a) thermal conductivity versus burnup (b) thermal conductivity versus pin radius.

(c) melt temperatures

• 15.11 The new TOP with failure to scram is now based on a single rod withdrawal ramp rate of 0.65t per second canpared to the earlier value which was 5.0c per second. Please explain this change.

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'. *15.12 Please' provide your estimates for the individual control rod and j control rod bank worth curves during .the various operational modes )

i in.cluding cold shutdown, hot standby, low power (e.g., 50% power) and j

full power operation. Describe these separately for the. primary and secondary rods. 1 l

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• 15.13 Please explain how you plan to justify basing the'. control rod withdrawal accidents on withdrawal of a single control rod and not a full rod. bank withdrawal.

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• 15.14 Please discuss the potential for failure of the bearing on the other-primary pump in the event of a primary pump seizure. . Also provide your 1 assessment of the potential consequences'if this should occur.  !

• 15.15 Please provide your updated analyses of the various Chapter 15 accident-studies presented at the October 15, 1987 meeting.

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. Request for'.Information on  ;

r SAFR Appendix A I

-l A.1 In Table A-1, the Fuel Axial expansion coefficient for (1) fuel '

unrestrained by cladding, (2) fuel restrained by cladding, anci i (3) fuel partially restrained by cladding should be provided.

4 A.2 Plots of reactivity versus time in Appendix A are on a long time scale, i.e., O to 800 seconds. In many cases, it is the first new seconds-that are the most important. Please provide plots' that better show the f first 30 seconds of the transients.

A.3 There is some ambiguity regarding the time of each of the. events described in the combined LOF/ TOP /LOHS analysis presented in Section-A.6. Do all three occur simultaneously? ' Also, isn't a LOHS with a j

delayed LOF quite probably a worse case, as the' natural circulation  !

head could be compromised by operation without heat removal through the IHX? i A.4 In the unscransned LOHS, the primary pumps continue to function. It I seems likely that a delayed trip of the primary pumps could aggravate the situation considerably, as the natural circulation head would be j reduced during the period after the IHTS is lost. Please explain the i 1

likely scenario once the IHTS is lost, i.e., at what time are signals j to (1) scram the reactor, and (2) trip the primary pumps, likely? Has

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RI looked at any transients in this case? I A5 The EBR-I reactor core was ultimately damaged by a bowing-related accident. Please explain the difference between the EBR-I, EBR-II, and SAFR reactors, and explain why the EBR-I accident is not relevant to the current technology.

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EnclosureL4'

,1 REQUEST FOR ADDITIONAL-INFORMATION FROM SAFR.0N WASTE DISPOSAL-AND RADIATION' PROTECTION'

.V 11.1 Section 11.3.2 " Design Description" of the PSIDl discusses .the options .

available.for processing, storing'and' disposing of metallic sodium and sodium bearing solids. It states that a firm' decision has not been made. regarding the method and. assigned responsibility for ' ultimate -

disposal At the operating! license-(OL).-stage,, the applicant will be required' L

to demonstrate that methods'for processing,J packdging., transporting and ultimate disposal of the waste have been developed that. willi satisfy the applicable criteria of 10 CFR.Part 61,10 CFR Part 71, 1

DOT Regulations, disposal site license conditions, Regulatory '

Guide 1.143 and Standard Review Plan.NUREG-0800, Section 11.4.

At this stage of the design the designer should commit.to developing  ;

waste processing, packaging, transportation and disposal' methods in accordance with these applicable _ criteria.

11.2 Describe your plans for compliance with items II.F.I', . Attachment 1,  !

Noble Gas Effluent Monitor,II.F.2, Attachment 2, Sampling and i i Analysis of Plant Effluents; and III.D.1.1, Integrity?of Systems OutsideContainmentLikelytoContainRadioactiveMaterial(as applicable to the SAFR) of NUREG-0737, " Clarification of TMI Action Plan Requirements". '

12.1 As specified in ~ Regulatory Guide 8.19,' you should use personnel exposure data for specific kinds of work and job functions 'available -

from similar operating plants. Describe infomation.you have obtained and plan-to use regarding source tems and occupational-radiation exposures for liquid metal reactors.

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.,. . ,e 12.2 As specified in Regulatory Guide.8.8, you should attain the-objectives;in Section:C to provide assurance-that exposures of station personnel. to' radiation will be> ALARA throughout the pl6nt,'

from planning and design' through decommissioning.. 'Is RI comitting -

to these objectives?

12.3- Is RI comitting.to the requirements'specified' in NUREG-0718. " Proposed )

' Licensing Requirements for Pending Ap'plications. for Construction Permits and Manufacturing License"-(II.B.2), as.they relate to.SAFR?-

12.4 Provide commitments to conform to'the. provisions of the followingl  !

Regulatory Guides, as they apply for SAFR, or describe alternative-measures to be taken to provide a comparable ' degree of worker. ]

protection.

1.97, " Instrumentation for Light-Water-Cooled. Nuclear Power Plants to Assess Plant and Environs Conditions During and-Following and Accident"  !

8.12. " Criticality Accident Alann Systems"'

8.13, " Instruction Concerning Prenatal Radiation Exposure" s

8.14. " Personnel-Neutron Dosimeters" 8.15. " Acceptable Programs for Respiratory Protection" 8.26, " Applications of' Bioassay for. Fission and Activation Products" 8.27, " Radiation Protection Training for Personnel at Light-Water Cooled Nuclear Power Plants"'

8.29, " Instruction Concerning Risks from Occupational Radiation-Exposure"

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' November 2;1987- i

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DISTRIBUTION.

RES Circ j

1 Chron ARGIB R/F E. Beckjord

.T. Speis i,

B. Morris Z..Rosztoczy

T. King .

J. N.' Wilson .

C.. Allen R. Landry, P. Williams J.. Flack' ,

M. Dey B. Hardin i R. Baer-- ,

N. Anderson 't F. Cherny j S. Shaukat 1 R. Johnson D. Thatcher J. Hulman J. Glynn 1 L. .Sof fer . .3 J. Read D. Cleary G. Arndt R. Kirkwood R. Erickson B. Mendleschn-E. Chelliah L. Beltracchi F. Congel L. Cunningham J. Minns i D. Matthews F. Kantor E. Podolak M. Spangler G. VanTuyle, Bhl M. El-Zeftawy, ACRS/H-1026

."PDR - Project 673 w/o enclosures-  ;

t ProjectJ11e? 673 E ( Chiya ljfj le s )lgp I

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