ML20148S016

Requests Comments on 880325 Meeting Re Safr PRA & Bounding Events by 880515.Attendee List & Meeting Summary Encl

ML20148S016
Person / Time
Issue date:	04/11/1988
From:	Morris B NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
To:	Nulton D ENERGY, DEPT. OF
References
PROJECT-673A NUDOCS 8804150231
Download: ML20148S016 (9)

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UNITED STATES o

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APR 111989 Mr. Davia hulton, Deputy Director Office of Advanced Reactor Programs Office of Nuclear Energy United States Department of Energy Washington. D.C.

20545

Dear Mr. Nulton:

i The NRC staff and its consultant (BNL) met with the DOE staff, contractor (RI) and consultant (ANL) on March 25, 1988 to discuss concerns regarding the SAFR (Project 673) PRA and Bounding Events.

Enclosures 1 and 2 provide the list of attendees and agenda, respectively, for the March 25 meeting with RI.

It is requested that RI document their responses to the questions resulting from this meeting as documented in Enclosure 3.

1he Bounding (Events (BEs) have been the subject of several discussions with both the GE PRISM) and R1 (SAFR) design teams during the last four months.

The latest list of BEs is attached as Enclosure 4.

In order to meet our Safety i

Evaluation Report schedule, any comments on these events should be provided by April 15, 1988. Analyses of the PRISM and SAFR conceptual designs for these events should be provided by May 15, 1988. The analyses should include assessments of the thyroid and whole body doses at the site boundary for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> and 30 days.

We also require an assessment, by May 15, 1988, of the enhanced Safety of these designs in accordance with Enclosure 5.

If there are further questions regardin any of the enclosures or dates, feel free to contact Dr. Ralph Landry (492-3735)g of my staff.

Sincerely, ZeYaa Bill M. Morris, Direc M '$

Division of Regulatory Applications Office of Nuclear Regulatory Research Enclosures As stated

't 8804100231 00M 11 PDR PHOJ PDR 673A g\\,

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SARF PRA MEETING March 25, 1988 Attendees R. Landry NRC/RES (301)492-3735 i

M. Dey NRC/RES (301)492-3730 J. Flack NRC/RES (301) 492-3741 T. King NRC/RES (301) 492-3765 J. Wilson NRC/RES (301) 492-3729 G. VanTuyle BNL (516)282-7960 T. Chu BNL (516) 282-2389 T. Ginsberg BNL (516) 751-6765 G. Slovik BNL (516)282-7983 J. R. Humphreys DOE (301)353-3588 i

G. Sherwood DOE (301)353-4162

0. Pedersen ANL (312)972-3335 R. Lancet R1 (818)718-3462 R. Amar R1 (818)718-3410 D. Rogers R1 (818)718-3410 P. Rutherford R1 (818)718-3393

,7.

SAFR PRA MEETING March 25, 1988 Agenda 8:30 am Introduction T. King (NRC)

R. Lancet (RI) 8:45 am Event Probabilities R. Lancet (RI) 11:15 am Energetics D.Pedersen(ANL) 12:00 Noon Lunch 1:00 pm Pump Questions D. Rogers (RI) 1:30 pm Combined Yessel Leaks R. Lancet (RI) 3:00 pm Summary R. Lancet (RI) 4:00 pm Adjourn

)

SAFR PRA Meeting March 25, 1988 Sumary A letter was sent on March 10, 1988 from B. Morris (NRC) to D. Nulton (00E)

The SAFR noting areas of concern which had arisen from the SAFR PRA review.

design team identified several questions in that letter and additional questions, The following or points of clarifications, in succeeding telephone conversations.

is the list of questions as discussed at the meeting.

i The following two events have been identified as potentially dominant 1.

contributions to the SAFR risk:

LOF with failure to scram compounded by failure of inherent shutdown capability.

Large earthquake which triggers pump trips while causing the control l

rods to become stuck thereby defeating both scram and the inherent feedbacks due to control rod driveline expansions.

Provide the bases for estimation of the probabilities of these events.

This assessment should include the probability of failure to scram and the conditional probabilities for consequential failure.

What is the probability that the second primary pump would fail soon 2.

after the first pump seizes?

For internal events analysis, what is the conditional probability of 3.

stuck rods given that a reactor trip signal has been generated?

(Substantiate quantifications with as much information and/or references as possible.)

O 4.

For internal events analysis and given stuck rods, what is the failure probability of the inherent reactivity feedback? (Substantiate quantification with as much information and/or references as possible.

5.

For the inherent reactor response, explain how f ailure rates and uncertainties were derived.

6.

In Figure 7.5 (SAFR PSID, Appendix C) there appears to be no basis for assignments of probabilities because the uncertainties are too large.

Explain the bases (including uncertainties) for the core vessel response.

7.

For seismic analysis, SAFR assumes a decreasing reliability of the inherent feedbacks for large earthquakes. Presumably, this is due to the increasing likelihood that the control rods would become stuck so that they could not expand freely into the core. Describe how these reliabilities were estimated given that these are risk-dominant sequences.

8.

For the decay heat removal system, explain how failure ratas and uncertair. ties were derived.

9.

With sodium boiling a 3-sec ramp rate can be achieved. Will the fuel move fast enough to counter this? (TREAT tests have not been conducted at this rate.)

10. Provide an assessnent of the accident scenario and consequences assuming a reactor vessel and guard vessel melt through.

11. Provide additional information regarding the following release assumption concerns:

The PRA assumed failure at the fuel melt temperature which leaves Ba and Sr in the fuel matrix. However, at higher temperatures, the Ba

l and Sr will be released into the sodium coolant. Explain why these higher temperatures are not possible or factor into your analysis the release of Ba and Sr when the higher temperatures are reached.

Some assumptions are based on oxide fuel where there is a lot of 02 available in the sodium to form compounds that hold up in the sodium.

This won't happen in the metal fuel, therefore, why are such assump' ions valid?

For large excursions large bubbles may be created which would transfer releases directly to the cover gas without going through the sodium.

How would such an assumption affect your release calculations?

12.

Is the SAFR plant protection system designed to delay pump trips until the reactor has been scrammed? If so, how does the system detect and ensure that the reactor has scrammed prior to tripping the pumps?

13. Under seismic events, can the control rods jam such that the thermal response doesn't work?

14 What is the likelihood of having one of the two primary pumps seize during the first minute of coastdown?

i

15. For large energetics will the reactor head assembly lift, expel sodium and fission products, and then reseat? What is the likelihood it will reseat?

I

16. Provide information regarding branch tree probabilities and uncertainties associated with those sequences which could lead to energetics events.

1 f

i

i Deteministic Events These events are intended to bound the LMR OBA and BDBA spectrum in order to account for PRA uncertainties and provide conservatism in selecting a SSST and assessing the adequacy of containment and offsite evacuation plans. These events are judged to be bounding for the following categories:

Reactivity insertion Heat removal Loss of coolant Na/H O reaction 2

Event-1 Inadvertent withdrawal of all control rods without scram for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (single module):

forced cooling j

RACS/RVACS cooling only Event-2 Station blackout for 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />.

loss of all ac power Event-3 Loss of forced cooling plus ORACS/RACS/RVACS with scram.

(single module):

25% partial unblockage after 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Event-4 Instantaneous loss of flow from one primary pump (single module);

coast down flow for other pump without scram for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Event-5 S.G. tube rupture with failure to isolate or dump water from S.G.:

justifiable number of tube failures defined sequence of ruptures Event-6 Large Na leaks (single module):

Double ended guillotine break of IHTS pipe RVleak(criticalleak)

Event-7 External events consistent with their treatment for LWRs.

Reouirements Associated with Enhanced Safety Applicant should assess and document enhanced safety f

characteristics / margins:

1 long response time reduced potential for operator error j

capability to retain FP highly reliable safety systems (passive / inherent l

characteristics) i sirplification(systems / analysis)

Potential improvements in safety are to be considered when the margins are small or when large improvements in safety can be realized with reasonable cost. These improvements could be selected for analysis and implemented using engineering judgement.

i Demonstrate enhanced safety / margins via testing.

4

9 David Nulton, Deputy Director DISTRIBUTION RES Circ Chron ARGIB R/F i

E. Beckjord T. Speis B. Morris Z. Rosztoczy T. King J. N. Wilson R. Landry P. Williams J. Flack B. Hardin

0. Gormley R. Baer N. Anderson F. Cherny S. Shaukat R. Johnson D. Thatcher J. Hulman J. Glynn L. Soffer J. Read D. Cleary G. Arndt R. Kirkwood R. Erickson B. Mendlesohn E. Chelliah L. Beltracchi F. Congel L. Cunningham

0. Lynch D. Matthews F. Kantor E. Podolak M. Spangler G. VanTuyle, BNL M. El-Zeftawy, ACRS/H-1026

-POR ' Project 673,.

Project File 673 (Central Files)