ML20057A803

From kanterella
Jump to navigation Jump to search
Responds to Staff Memo Re Briefing in Progress of Design Certification Review & Implementation on 930602
ML20057A803
Person / Time
Issue date: 09/09/1993
From: Taylor J
NRC OFFICE OF ENFORCEMENT (OE)
To: Rogers, Selin I, The Chairman
NRC COMMISSION (OCM)
References
REF-10CFR9.7 M930602A, NUDOCS 9309150341
Download: ML20057A803 (13)
v • d • e


Text

.

g m_.,

O k

UNITED STATES

(

/

NUCGAR REGULATORY COMMISSION P

WA5HINGTON, D.C. 20555 4001 September 9, 1993 MEMORANDUM FOR:

The Chairman Commissioner Rogers Commissioner Remick Commissioner de Planque FROM:

James M. Taylor Executive Director for Operations

SUBJECT:

STAFF REQUIREMENTS - BRIEFING ON PROGRESS OF DESIGN CERTIFICATION REVIEW AND IMPLEMENTATION, JUNE 2, 1993 (M930602A)

This memorandum responds to the staff requirements memorandum (SRM) of June 24, 1993, in which the Commission asked for additional information on the status of the staff's design certification review of standardized light water reactor designs. A previous memorandum responded to some of the issues in the SRM. This memorandum is intended to reply to the request that the staff provide a description of the code validation and data usage planned for the passive plant testing programs.

The passive reactor certification and test plan will provide the staff and reactor vendors with experimental data for evaluating the ability of the analytical computer codes to predict the behavior of the passive reactor designs. The plan, which will proceed according to the schedule indicated in the Enclosure, includes separate effects tests and integral tests. These tests will provide data for the assessment of the codes to model particular phenomena and assess the performance of the codes as a whole during complex transients. The staff and the vendors will perform calculation > of selected experiments and will compare their calculations to the data.

Such comparison will allow staff and vendors to understand the strengths and weaknesses of the codes, to feed this information back into the code development process, and to try to quantify the degree of uncertainty in the calculations.

Vendor Codes The staff's review of the vendor code qualification information submitted for the passive plants is in the very early stages. Although the staff has received some information from both passive plant vendors, the information received is not sufficient to complete the review, and the staff has requested additional supporting documentation.

Both vendors have performed initial assessments of the phenomena and para-metric ranges applicable to the passive plants and have developed their test programs to provide data in the areas for which, in their judgment, there is i

not an adequate data base for analytical model development and validation.

The staff is currently engaged in reviewing the test programs, an effort that is expected to continue into FY 1995. The testing review is being coordinated closely with the computer code review, reflecting the dependence of the code m3;

/y 1

15r068

g8998 = = vov g

INICRE EdTWiGfdW j PT9.7 PDR DI

The statf has developed a detailed plan for reviewing and monitoring the i

vendors' test programs. The plan includes reviews of tne test facility i

designs, scaling bases, in:trementation, and test matrices, ;o ensure that the test facilities and planned experimental programs will be able to provide the necessary data for design certification. Staff members will visit the facili-ties during selected tests to monitor the performance of the test operators and test equipment. After the test programs are completed, the staff will review the test data and analyses to determine the adequacy of the data and their applicability to the passive plants. The staff's review of each vendor's design certification test programs will be incorporated into the respective safety evaluation report (SED.) for each passive plant.

For Westinghouse's AP600, three computer codes are ceing used to perform reactor coolant system design-basis accident asressments: WCOBRA/ TRAC, a best-estimate code for LOCAs; NOTRUMP, an " Appendix K" conservative evaluation model for small-break LOCAs; and LOFTRAN, a conservative code for non-LOCA transients. The first part of the code qualification documer tation for WCOBRA/ TRAC has been submitted to the NRC.

l The staff expects to base validation of the AP600 codes on existing experi-ments and plant data, where such information can be shown to cover parametric ranges similar to those in the AP600, and on data developed in the AP600 testing program for phenomena and parametric ranges that are unique to the passive design. Key experiments in the AP600 test program will not be completed until early-to-mid 1994, and data analysis reports for these test programs are not expected until later in 1994.

For General Electric's (GE's) simplified boiling-water reactor (SBWR), code qualification documentation has been received for TRACG, a best-estimate code being used for design-basis accident assessment. GE has provided an outline of the qualification basis for TRACG, which is similar to that described above for the AP600 codes. Tha existing BWR plant and experimental data base is used, as well as the test data from the SBWR test programs for phenomena and parameters specific to the SBWR.

GE's design certification test program began in 1988 and is expected to continue until 1995.

For those areas in which information is available for completed test programs, detailed assessments of the test facility designs, test matrices, data, and analyses are in progress. As information is provided to the staff on planned testing, similar evaluations will be performed, in the j

manner described above.

The staff will proceed with its reviews of the vendor's containment codes in a similar fashion to assess the ability of the codes to predict the temperatures and pressures inside containment for a variety of accidents, as well as the

i behavior of hydrogen and other non-condensable gases. The staff is also con-I sidering the ability of its own best-estimate containment codes to establish pressure-temperatuare profiles for conformance to 10 CFR 50.49 criteria (equipment qualification), and it will evaluate the relative performance of established codes such as CONTEMPT LT-28 and CONTAIN for this purpose.

NRC Codes The staff is using similar methods to evaluate its own computer codes. NRC staff and contractors are using the coupled RELAP5/CONTAIN code to analyze safety system performance in both the AP600 and SBWR. Assessment of RELAPS will be performed against test data from both the applicant testing programs and the NRC confirmatory testing programs at ROSA-V for the AP600 and at Purdue for the SBWR. The staff is focussing its research program on those transients that rely on the passive safety injection systems.

In order to evaluate code accuracy and uncertainty, phenomena identification and ranking tables (PIRT) are being developed for the AP600 and SBWR with the assistance of the Commission-approved group of thermal-hydraulic consultants. The PIRT will be used to help identify the important phenomena that the codes must model.

It also will be used to guide the evaluation of code vs. data compari-sons, including defining the plant sensitivity studies needed to arrive at a determination of code accuracy and uncertainty. The code scalability accuracy and uncertainty (CSAU) process, developed and applied previously on NRC-developed large system codes (e.g., TRAC), will be followed in assessing the code accuracy and uncertainty, with modification to conform with the actual code improvement and assessment being used for the advanced designs, j

0 tginal signed by Ames IA. Taylor l

James M. Taylor Executive Director for Operations

Enclosure:

I Advance Reactor Testing Outline cc:

SECY d

OGC W

OCA

~

&,o 4 -

,, 3b {,2 OPA y\\u\\ h~lM < '

>V/

l Tech Ed: Dee - 8/20 3 o\\

,/3j yb ML" D:DSSA"(Y L:SAS':DSSA y

.SRXB B.

STHE A

RES [ in hU G

RCafbsoml R ones R af 'ett DMc r/on LShotk AThadanib 8/li)/93 8

/93 8/24/93 8/,f///93 8/g/93 8/gi/93 ED[

ADT437 D, 'k kfield f

WTRussell Ti u y

JTa lor 8/30/93 df/

/93

$/

/93

/93 Document Name: A:\\ WIT 102.RC S

f./ /

)

_ 4-DISTRIBUTION.,

LCentral - files DSSA Rdg i

-PDR s

JTaylor TMurley FMiraglia WRussell JPartlow TGody DCrutchfirld RCaruso Caruso CHRON AThadani l

RJones RBarrett GMcPherson LShotkin BToms NRR Mailroom - WITS 9300102 ED0 9300102 EDO R/F l

JSniezek j

HThompson l

JB1aha 0GC OPA OCA RBorchardt TEssig i

RHasselberg PShea SGreen PMannanelli j

JWilson l

MMalloy i

1 l

w P

w ir e--

w w

a m-

  • i ENCLOSURE

~

O e

i t

i i

r t

ADVANCED REACTOR TESTING OUTLINE t

I I

i I

I t

P e

I i

i r

l

)

i i

- l l

i

AP600 CERTIFICATION AND TEST SCHEDULE

~

IIIIIIIIIII IIIIIIIIIII IIIIIIIIIIi 1993 1994 1995 Staff W RAI W DSER RAls Resp.

Staff DSER

Response

Staff FSER A

Cntmnt Test

\\'

8 A

CMT Tests SPES-2 4

A Tests 8

A OSU Tests t

A ADS Tests ROSA-5 Tests l IIIIIIIIII IIIIIIIIIII IIIIIIIIIII 1993 1994 1995 SER schedule reflects SECY-93-097 8 = Denotes submittal of Quick Look reports Denotes beginning of submittal of processed data to the NRC A =

Shid6d Areas = Denote time from test completion to final report to the NRC r-

~ --.-----w- -----

AP600 REVIEW AND TESTING SCHEDULES

__ __ __ __ __ __ __ __ __:__:__!__ __:__ __ __:__ __ __i__i__ __ __:__i__i__i__i__ __ -_i__i__ __i__ __i__

J F M A M J J A S O N D J F M A M J J A S O N D J T M A M J J A S O N D 1993 1994 1995 LICENSING SCHEDULE t

t t

t i

DSER DSER FSER FSER FSER PM COMM PM COMM ISS SEPARATE EFFECTft TESTS oooooooooo Testing in progress t

Issue final report PCCS Heat Transfer--1/8-Scale Phase 1 Tests 1992 Phase 2 Tests ocoooooooooo

?

PCCS Water Distribution Phase 1 Tests 1991 Phase 2 Tests 1991-2 Phase 3 Tests ooooooo

?

PCCS Wind Tunnel Phase 1 Tests 1991 Phase 2 Tests 1992 Phase 3 Tests TBD Phase 4 Tests oooooooooooo t

PRHR HX Phase 1 1989, Report 1990 Phase 2 1990, Report 1992 ADS Phase A (Sparger) 1992 i

Phase B (Valves) ocooooooooo t CMT oooooooooo 1

- - ~...

AP600 REVIEW AND TESTING SCHEDULES

__g__!__;__3__g__! _l__3 _!__ -_!__g__ -_l__;__1__l__ -_:__!__!__ -_!__l__!__g__;-_g--l__g__l__!__1__g__l__;

J F M A M J J A S O N D J F M A M J J A S O N D J F M A M J J A S O N D 1993 1994 1995 1

LICENSING SCHEDULE t

i f

f f

DSER DSER FSER FSER FSER PM COMM PM COMM ISS INTEGRAL SYSTEMS TESTS oooooooooo Testing in progress t

Issue final report SPES-2 oooooooooooooooooooooo i

OSU 0000000000o3000o00000.

t NRC CONFIRMATORY INTEGRAL TESTS ROSA V Shakedown oooooo Initial Matrix Tests (Phase 1) ooooooo ooooooooooooooo Additional Matrix Tests (Phase 1) xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx Phase 2 Tests (Possible)

.. ~

SBWR REVIEW AND TESTING SCHEDULES

__i__:__ __:__ __ __!__:__:__: _:__ __!__!__!__!__:__ __:__!__:__ __!__ __ __!__:-_i__ __ __ __ __ __:__ __

A M J J A S O N D J F M A M J J A S O N D J F M A M J J A S O N D J F M 1993 1994 1995 1996 LICENSING SCHEDULE t

t t

t i

DSER DSER FSER FSER FSER PM COMM PM COMM ISS SEPARATE EFFECTS TESTS

  • oooooooooo Testing in progress t

Issue final report MIT Condensation Heat Transfer Tests **

Steam _ Air Tests 1990 Steen-Air / Steam-Helium Tests 1991 UC Derkeley Condensation Heat Transfer Tests **

NC Steam-Air Tests 1990 NC and FC Steam-Air / Steam _ Air _ Helium Tests 1991 Additional Tests ooooooo?????????

PANTHERS PCCS Tests ococooooooooooooooooooooooo t

PANTHERS IC Tests ooooooooooooooooooooooo t

INTEGRAL SYSTEMS TESTS

  • GIRAFFE Phase 1 1990**

GIRAFFE Phase 2 1992**

DIST (CDCS) Tests 1988 Report 1989 PANDA oooooooooooooooooooooooooooooooooooooooooooooooooooooo t.

NRC CONFIRMATORY TESTING oooooooooooooooo-__>t1997 I

  • Test schedules are preliminary and subject to change
    • Final report not yet issued, date uncertain

]

l i

8 AP600 TEST LOGIC l

Reactor Systems Separate Effects Tests Integral Systems Tests i

  • SPES-2
  • OSU
  • CMT
  • PRHR Heat Exchanger Tests NOTRUMP, WCOBRA/ TRAC, LOFTRAN Code Development Tested Future ECCS I

Iterations WGOTHIC m

code DM Contelnment Containment Separate Effects Tests IntegralTest e PCCS HeatTransfer

  • Large Scale PCCS Tests
  • PCCS Water Distrib.
  • PCCS Wind Tunnel i

I

4 I

s-i SBWR TEST LOGIC

^

Separate Effects Tests Integral Systems Tests e GIST e MIT/UCB Condensation

, ffND Tests

  • PAhTHERSIC PCCS Tests Tests litAC G Code Development Tested Future ECCS Meradons Codes Code Development

AP600 KEY PHENOMENA I

PHENOMENA Gravity drain ECC injection Condensation during ECC injection /recirc CMT TESTS Recirculatory flow during ECC injection Thermal stratification SEPARATE Depressurization effects on CMT injection EFFECTS Natural convection heat transfer in PRHR TESTS tubes TESTS Boiling heat transfer in IRWST Critical flow during depressurization ADS TESTS Condensation in IRWST during depressurization Condensation with non-condensibles Condensate flow on containment inner surface PCCS TESTS Containment outer shell cooling with and without water flow Water distribution on containment outer shell Wind effect on containment cooling High-pressure system response to design basis events INTEGRAL Safety and non-safety systems SPES-2 interaction SYSTEMS Natural circulation loop flow Integration of independent TESTS component /phenomenological models Long-term cooling behavior OSU Effect of containment backpressure Low-pressure system response and system interactions

i SBWR KEY PHENOMENA i

I PHENOMENA Heat transfer with noncondensibles in MIT/UCB TESTS natural and forced convection (steam / air and steam / air / helium)

SEPARATE Heat exchanger performance EFFECTS Flow distribution through tubes TESTS PANTHERS Natural / mixed convection heat transfer in tubes l

HX TESTS Nattral convection / boiling heat transfer in pool Performance of non-condensibles venting system Recirculatory cooling greater than one GIRAFFE hour post-accident TESTS Performance of non-condensibles venting system PCCS heat transfer behavior INTEGRAL GIST TESTS Gravity drain ECC injection SYSTEMS Recirculatory cooling greater than one hour post-accident TESTS Performance of IC and PCCS heat PANDA exchangers TESTS Performance of non-condensib1 s venting system j

Steam /non-condensibles distributions in drywell post-LOCA Systems interactions (limited) i Integration of l

component /phenomenological models i

1 1

Retrieved from "https://kanterella.com/w/index.php?title=ML20057A803&oldid=4479993"