ML20214U707

Summary of 861120 Meeting W/Util,Ornl,Ga Technologies & Eg&G Re Temp Profiles for Equipment Qualification.List of Attendees & Viewgraphs Encl

ML20214U707
Person / Time
Site:	Fort Saint Vrain
Issue date:	12/02/1986
From:	Heitner K Office of Nuclear Reactor Regulation
To:	Berkow H Office of Nuclear Reactor Regulation
References
TAC-42527, NUDOCS 8612090364
Download: ML20214U707 (27)

Text

_. . .--

[f, s A' '.[8 p([$ g UNITED STATES NUCLEAR REGULATORY COMMISSION l

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$ / *: j WASHING TON, D. C. 20555

% , , [, , *# December 2, 1986 Docket No. 50-267 MEMORANDUM FOR: Herbert N. Berkow, Director Standardization and Special Projects Directorate Division of PWR Licensing-8, NRR THRU: Oliver D. T. Lynch, Jr., Section der Standardization and Special Projects Directorate l Division of PWR Licensing-B, NRR FROM: Kenneth L. Heitner, Project Manager Standardization and Special Projects Directorate Division of PWR Licensing-B, NRR

SUMMARY

OF NOVEMBER 20, 1986 MEETING WITH PUBLIC SERVICE COMPANY OF COLORADO (PSC) TO DISCUSS FORT ST. VRAIN (FSV)

TEMPERATURE PROFILES FOR EQUIPMENT QUALIFICATION This meeting was requested by PSC in response to the October 30, 1986 letter from the NRC to PSC concerning FSV temperature profiles for equipment qualification. In this letter, the NRC had concluded that the calculational methods used.by PSC were not conservative. The purpose of this meeting was to review PSC's calculations and methodologies, and find a potential basis for NRC approval of the FSV temperature profiles. Attendees at this meeting are listed in Enclosure 1.

PSC presented a sumary of their basis for concluding that their tiemperature profile calculations were conservative (se'e Enclosure 2). PSC stated that the different computer codes used by the staff's consultant and PSC were

not the cause of the different temperature profiles. However, the lower heat transfer coefficients used by the staff's consultant were the primary l cause of the differences. PSC noted the higher heat transfer coefficients that they had used were derived from original licensing calculations, and were accepted by the staff when the plant was originally licensed. PSC felt these heat transfer coefficients were still conservative.

PSC had investigated the possibility of taking credit for potential limita-tions in plant power level. However, preliminary calculations for scenario HRH-2 at 75 percent of full power still exceeded the design temperature profile.

8612090364 861202 PDR ADOCK 05000267 P PDR

PSC further noted that large conservatisms existed in the restricted volumes they had used for their evaluations in the reactor and turbine buildings.

The actual building volumes were considerably higher. However, preliminary calculations with less conservative building volumes gave much lower tempera-ture profiles. PSC maintained that the temperature profiles used for equipment qualification could not easily be modified, since most of the equipment qualification files were complete. For certain critical components, such as cables, the margin in temperature for equipment qualification was small.

The staff held a separate caucus. The staff then directed PSC to redo the sample temperature profiles with the following guidance:

Use heat transfer coefficients similar in value to the staff's consultants (approximately one in English units).

Use revised building volume and surface area parameters to reflect additional volumes where steam could easily penetrate, and Do the sample calculations for a large break (i.e., HRH-2), small break (i.e.,CRH-19),andoperator-isolatedbreak. Three revised calculations should be done for each building.

The staff also recommended that PSC provide a separate evaluation of other i

factors which provided additional conservatisms in these calculations, such as the blowdown orifice coefficient and thermal radiation from the hot gas i to colder surfaces.

PSC stated that additional time would be needed to perform these new analyses. They estimated that analyses for the first" building could be submitted by mid-December 1986, and for the second building by the end of l December 1986.

The staff noted that this could potentially impact the schedule for the field inspection and would advise PSC if a schedule change was needed.

(Subsequent to the meeting, the staff advi' sed PSC that the current schedule for the field inspection would be maintained).

h Kenneth L. Heitner, Project Manager Standardization and Special Projects Directorate Division of PWR Licensing-8, NRR

Enclosures:

As stated cc w/ enclosures:

See next page I

- - - - , - - ,,,- - - - . - - c - - , - - - , - - . - - - - - - - - . - - . ,

PSC further noted that large conservatisms existed in the restricted volumes they had used for their evaluations in the reactor and turbine buildings.

The actual building volumes were considerably higher. However, preliminary calculations with less conservative building volumes gave much lower tempera-ture profiles. PSC maintained that the temperature profiles used for equipment qualification could not easily be modified, since most of the equipment qualification files were complete. For certain critical components, such as cables, the margin in temperature for equipment qualification was small.

The staff held a separate caucus. The staff then directed PSC to redo the sample temperature profiles with the following guidance:

Use heat transfer coefficients similar in value to the staff's consultants (approximately one in English units),

Use revised building volune and surface area parameters to reflect additional volumes where steam could easily penetrate, and Do the sample calculations for a large break (i.e., HRH-2), small break (i.e., CRH-19), and operator isolated break. Three revised calculations should be done for each building.

The staff also recommended that PSC provide a separate evaluation of other factors which provided additional conservatisms in these calculations, such as the blowdown orifice coefficient and thermal radiation from the hot gas to colder surfaces.

PSC stated that additional time would be needed to perform these new analyses. They estimated that analyses for the first building could be submitted by mid December 1986, and for the second building by the end of December 1986.

The stoff noted that this could potentially impact the schedule for the i

field inspection and would advise PSC if a schedule change was needed.

l (Subsequent to the meeting, the staff advised PSC that the current schedule for the field inspection would be maintained).

1 Kenneth L. Heitner, Project Manager l Standardization and Special Projects Directorate Division of PWR Licensing-B, NRR

Enclosures:

As stated DISTRIBUTION:

cc w/ enclosures: Docket File KHeitner See next page NRC PDR OGC-Bethesda Local PDR EJordan PBSS Reading BGrimes JPartlow ACRS (10)

. HBerkow NRC Patficipants PBSS h P PBS9p oonan KHeitner:ac OL n h HBenRow 110(g86 11/ l /86 11/l /86 IJ/>/86

\

g Enclosure 1 FSV Temperature Profile Meeting LIST OF ATTENDEES Name Affiliation J. C. Conklin Oak Ridge National Laboratory Charles Hinson NRC/NRR/DPLB/PBSS Robert Jones NRC/NRR/DPLB/RSB J. S. Wermiel NRC/NRR/DPLB/PEICB Paul Shemanski NRC/NRR/DPLB/PEICB Jose A. Calvo NRC/NRR/DPLB/PEICS Richard E. Ireland NRC/RIV/RSB-ES K. L. Heitner NRC/NRR/DPLB/PBSS Mark White Battelle/PNL Carl Wheeler PNL Don Warembourg PSC M. L. Holmes PSC/Nucl. Licensing M. E. Niehoff PSC/Nucl. Design Mgr.

Carmels Rodriguez GA Technologies-Engineering F. W. Tilson PSC/Nucl. Engineering Ken Dvorak PSC/Nucl. Engineering Art Barsell GA Technologies F. C. Dahms GA Technologies John C. McKinley NRC/ACRS Staff Julio Landoni GA Technologies Max Yost EG&G, Idaho Norman Wagner NRC/NRR/DPLB/PEICSB John Ridgely NRC/NRR/DPLB/RSB

0. D. T. Lynch, Jr. NRC/NRR/DPLB/PBSS Herb Berkow NRC/NRR/DPLB/PBSS Dennis Crutchfield NRC/NRR/DPLB Chang Li NRC/NRR/PWR-A/PSB i

i l

1

ENCLOSURE 2 l

r 3 1

d DISCUSSION ITEMS

O THE ISSUE O HRH-2 BLOADOAN ANALYSIS

O COMPARISON OF CVTR TEST DATA

TO FSV ANALYSIS O LITERATURE SUPPORTING PREVAILING HEAT TRANSFER COEFFICIENTS O CONSERVATISMS IN ANALYSIS

't )

[CALE']51 12 18-HOV-86

F

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THE ISSUE: APPROPRIATE VALUES FOR i HEAT TRANSFER COEFFICIENTS FOR FORT ST. VRAIN STEAM UNE BREAKS 1

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[CALE)51 11 18-HOV-86

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CO \ SBVA-~ S V S o THE UCHlDA HEAT TRANSFER COEFFICIENT UTILIZED FOR THE CONDENSING REGIME OF THE FSV BLOW- 3q i

DOWN ARE RECOGNIZED AS CONSERVATIVE.

o THE CONVECTIVE HEAT TRANSFER COEFFICIENT IS ' '

CONSERVATIVE BASED ON ACTUAL TEST DATA.

o TOTAL HEAT REMOVAL IS LOW FROM GAS MIXTURE  ;

(TAGAMI CORRELATION)

o BOUNDING BLOWDOWN ORIFICE COEFFICIENT (1.0) g.

l lS CONSERVATIVE.

o VOLUMES AND HEAT SINKS UTILIZED IN DEVELOPMENT OF FSV TEMPERATURE PROFILES O ,

ARE CONSERVATIVELY LOW.

1

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i HRH-2 BLOWDOWN ANALYSIS l

l O DURING BLOADOAN i

t H CONv VARIES LINEARLY FROM 5 AT

-START TO 7.5 BTU /FT*-H- F AT END OF BLOADOWN H COND IS BASED ON UCHIDA i

- 2 H TOTAL IS ABOUT 15 BTU /FT -H- F AT END OF BLOADOAN I

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[CALE]51 9 18-NOV-86 i

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( 3

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HRH-2 BLOWDOWN ANALYSIS (CONT) i

! O AFTER BLOWDOWN

- 2 H CONV IS EQUAL TO 5 BTU /FT l -H- F l t H CONYo IS BASED ON UCHIDA

[CALI]S1 10 18- NOV-86

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l REACTOR BUILDING TEMPERATURE RESPONSE TO -

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PIG.1 CVTR CONTAINM ENT STRUCTURC, 9 ,

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j COMPARISON OF CVTR TEST 3 WITH FSV HOT REHEAT HEADER

! BLOWDOWN (REACTOR BUILDING) - HRH-2 1

CVTR j PARAMETER TEST 3 HRH-2 BLOADOAN TIhE, SEC.

) 170 13 TOTAL STEAM ENERGY RELEASE, BTU 1.9 x 10 7 1.2 x 10 7

{ ANALYSIS VOLtNE, FT 3 227,000 534,730 HEAT SIM( AREA, FT 2 37,120 282,550 l '

STEAM RELEASE RATE AT 10 SEC, LB/H 3.2 x 10 3 2.3 x 10 s i

SPECIFIC ENTHALPY AT 10 SEC, BTU /LB 1264 1535 TOTAL STEAM RELEASE, LB IISx10 4 7.9 x 10 3

PEAK GAS TEhPERATURE, *F 209 371 Y

fcAttls s is-Nov-es

j . .

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COMPARISON OF CVTR TEST DATA WITH CONTEMPT-G l MODEL RESULTS l , TOTAL VOLUME HEAT TRANSFER COEFF. PEAK PEAK 2

BTU /FT -H- F PRESSURE TEMPERATURE

! CASE CONVECTION C0f0ENSATl0N PSIG *F CONTEWT-G MODEL RUNS 1 2 UCHIDA 27.3 361 2 5 UCHIDA 23.5 281 3 10 UCHIDA 20.9 224

, 4 20 UCHIDA 19.4 215 5 20 2 x UCHlDA 16.5 209 l

l CVTR TEST 3 DATA + 18.0 209 3

! DURING BL0nOCMN

+ TEST 4 AND 5 CONDITIONS AND BLOADOWN RESULTS WERE SIMILAR i

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'350 , , , ,

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g o i END OF BLOWDOWN * ..

3 . t . . 3 I I 3 3 8 i  !  !  !  ! ' .: 3 50 , , , , , , ,

0.0 50.0 10 0.0 15 0.0 200.0 250.0 300.0 350.0 400.0 TIME - SECONDS ST9760 11/13/86 bconv = 20/5 DURING/AFTER BLOWDOWN; UCHIDA CONDENSATION

4 I 3 UTERATURE SUPPORTING PREVAIUNG HEAT TRANSFER COEFFICIENTS BY CONVECTION (HTCV) DURING STEAM j , BLOWDOWNS O

SLAUGHT[RBECK'S RECOWEiOATION FOR AN INITI AL HTCV IS 5

)

BTU /(FT -H "F)

O BENHAM, ET AL, ADOPTED SLAUGHTERBECK'S RECOWENDATION FDR PREVIDUS FSV EO %ORK i

j O ALMENAS'S METHODOLOGY BASED ON MATCHING CVTR, LEADS TO 2

! HTCV OF ABOUT 10 BTU /(FT -H- F)

O FRANK, ET AL, METHODOLOGY, BASED ON MATCHING CVTR, 2

GENERATES AN ALL CONVECTIVE HTCV OF 20 BTU /(FT -H- F)

O JUBB'S METHODOLOGY BASED ON BOILER LEAK TESTS LEADS TO 2

AN HTCV OF 5 BTU /(FT -H- F)

L ) ,

[CALE]51 0 18-NOV-86

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CONSERVATISMS IN ANALYSIS l

j 1. RELATIVE TO CVTR, CONVECTIVE j HEAT TRANSFER COEFFICIENT IS CONSERVATIVE BY FACTOR OF 3.

[CALE)51 4 18-NOV-86  ;

I

(

3 i

i -

CONSERVATISMS IN ANALYSIS (CONT)

2. RELATIVE TO TAGAMl, HEAT i

REMOVAL FROM ATMOSPHERE IS l

CONSERVATIVE BY A FACTOR OF 4.

i l

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3 , . . , , . , , .

i

t .

END OF BLOWDOWN t

SWE GA hhnologies M I 12 -

i l l l l

10 -

HEAT REMOVAL COMPARISON BETWEEN THE i . l 1

o HRH-2 SCENARIO! AND TAGAMI WITH THE TEMPERATURE PROFIT.E DEVELOPED FOR HRH-2 l i

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2 4 8 8 10 12 14

~ TIME, 8

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[ ,

i J CONSERVATISMS IN ANALYSIS (CONT) l

3.

ALL BLOADCWNS ASSUMED TO EXlT l

' PIPES WITH AN EFFECTIVE FLON j

ORIFICE COEFFICIENT OF 1.0.

)

(cattist s in - no v.. n c

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m i

l CONSERVATISMS IN ANALYSIS (CONT) i 4. VOLUMES AND HEAT SINK AREAS UTILIZED IN THE CONTEMPT-G CALCULATIONS ARE CONSERVATIVELY LOW.

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[ CALF j51 2 18 - plov- n r,

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"%$2## ~RuC10R eu,Loi~0 REACTOR VESSEL ~

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E a =LUMES INCLUDED = L 1mS k Ly FUEL STORAGE AND

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ANALYSIS er,";<-

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k VOLUMES INCLUDED IN ANALYSIS 0F TEMPERATURE PROFILES.

93 SM ADDITIONAL VOLUMES EXPOSED TO STEAM NOT INCLUDED IN ANALYSIS.'

REFUE ING ' ,

PESTRESSED N -

N CORE KAC SSEL % ~

STEAM GENERATOR MODULE % k k -

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f AUXILIARY .: 7 CONDENSER AUXILIARY OR ' ~

p BAY AEA Cl

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\ TUR81NE BUILDING TRA FORMER

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REACTOR BUILDING  !

o %

1 8

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1; Figure 1.2-3 Section Through Reactor Building and Turbine Building

[

REACTOR BUILDING SURFACE AREAS AND VOLUMES Area Volume fta ft3

• Around PCRV 282,550 534,730 Region east of 4A wall 157,470 327,310 Above Refueling floor 68,420 711,160 TURBINE BUILDING i

SURFACE AREAS AND VOLUMES Area Volume fta ft3 Below Operating floor 296,990 750,000 Above Operating floor 68,780 920,000 Aux. Bay Area Not calculated Not calculated E

i l Used in CONTEMPT-G profile calculation l

1 i

i 4

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4 CO\C _L S O\ S

1. HEAT TRANSFER COEFFICIENTS .USED ARE. CONSERVATIVE.

2. ANALYTICAL APPROACH HAS OTHER MAJOR CONSERVATISMS IN ADDITION TO HTC CONSERVATISMS.

3. COMPOSITE TEMPERATURE PROFILE IS CONSERVATIVE AND APPROPRIATE FOR FORT ST. VRAIN EQUIPMENT QUALIFICATION.

^

e_ _ __ . _

• Mr. R. O. Williams Public Service Company of Colorado Fort St. Vrain .

cc:

Mr. D. W. Warembourg, Manager Albert J. Hazle, Director Nuclear Engineering Division Radiation Control Division Public Service Company Department of Health of Colorado 4210 East lith Avenue P. O. Box 840 Denver, Colorado 80220 Denver, Colorado 80201 Mr. David Alberstein, 14/159A Mr. J. W. Gahm, Manager GA Technologies, Inc. Nuclear Production Division Post Office Box 85608 Public Service Company of Colorado San Diego, California 92138 16805 Weld County Road 19-1/2 Platteville, Colorado 80651 Mr. H. L. Brey, Manager Nuclear Licensing and Fuel Division Mr. P. F. Tomlinson, Manager Public Service Company of Colorado Quality Assurance Division P. O. Box 840 Public Service Company of Colorado Denver, Colorado 80201 16805 Weld County Road 19-1/2 Platteville, Colorado 80651 Senior Resident Inspector U.S. Nuclear Regulatory Commission Mr. R. F. Walker P. 0. Box 840 Public Service Company of Colorado Platteville, Colorado 80651 Post Office Box 840 Denver, Colorado 92138 Kelley, Stansfield & 0'Donnell Public Service Company Building Commitment Control Program Room 900 Coordinator 550 15th Street Public Service Company of Colorado Denver, Colorado 80202 2420 W. 26th Ave. Suite 100-D Denver, Colorado 80211 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 ,

Chairman, Board of County Commissioners of Weld County, Colorado Greeley, Colorado 80631 Regional Representative Radiation Programs Environmental Protection Agency 1 Denver Place 999 18th Street, Suite 1300 Denver, Colorado 80202-2413

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