ML20008E747

Forwards Evaluation of Cladding,Swelling & Rupture Models for LOCA Analysis,Per NUREG-0630,in Response to NRC 801014 Ltr.Fsar Will Be Amended by 810501 to Reflect Ten Grid Fuel Design

ML20008E747
Person / Time
Site:	South Texas
Issue date:	02/27/1981
From:	Goldberg J HOUSTON LIGHTING & POWER CO.
To:	Eisenhut D Office of Nuclear Reactor Regulation
References
RTR-NUREG-0630, RTR-NUREG-630 ST-HL-AE-623, NUDOCS 8103090558
Download: ML20008E747 (6)

Text

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The Light company a_,ee i.ix ies & re e,s no. nes ivoo iiees,ee.1<xms 77ooi <7is> 22 -92ii February 27, 1981 ST-HL-AE-623 SFN: C-0100 V-0100 9

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Mr. Darrell G. Eisenhut Division of Project Management 71 (fhMQs\

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Nuclear Regulatory Commission Washington, D. C. 20555 j 4' p 3 ./ $

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Dear Mr. Eisenhut:

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,,E' South Texas Project Units 1&2 Docket Nos. STN 50-498, STN 50-499 Cladding Swelling and Rupture Models for LOCA Analysis On October 14, 1980, Houston Lighting & Power Company received a letter from your office requesting additional infomation concerning the application of the cladding swelling and rupture models for Loss of Coolant Accident (LOCA) analysis. Specifically it was requested that HL&P provide supplemental infoma-tion which utilizes the materials models of drift NUREG-0630. In response to the above mantioned request, attached is the valuation of the potential of using fuel rod models presented in NUREG-0630 on the LOCA analysis' for the South Texas Project, Units 1&2.

This evaluation is based on a ten (10) grid fuel design. The South Texas Project (STP) FSAR currently reflects the nine (9) grid fuel design and the STP-FSAR will be amended by May 1,1981 to reflect the ten grid fuel design.

If there are any questions concerning this item, please contact Mr. Michael E. Powell at (713) 676-8592.

Very truly yours, M QJ. H.DGoldberg

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Ii Vice President Nuclear Engineering & Construction MEP/ par Attachment i

810809 0 %E A -

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flouston 1.ighting & Pemer company February 27, 1981 cc: J. H. Goldberg ST-HL-AE-623 D. G. Barker SFN: C-0100 Howard Pyle V-0100 R. L. Waldrop Page 2 H. R. Dean D. R. Beeth J. D. Parsons G. B. Painter L. K. English J. W. Briskin '

R. A. Frazar  !

H. S. Phillips (NRC)

J. O. Read (Read-Poland,Inc.)

M. D. Schwarz (Baker & Botts)

R. Gordon Gooch (Baker & Botts)

J. R. Newman (Lowenstein, Newman, Reis & Axelrad)

Director, Office of Inspection & Enforcement Nuclear Regulatory Comission Washington, D. C. 20555 M. L. Borchelt Charles Bechoefer, Esquire Executive Vice President Chairman, Atomic Safety & Licensing Board Central Power & Light Company U. S. Nuclear Regulatory Comission P. O. Box 2121 Washington, D. C. 20555 Corpus Christi, Texas 78403 R. L. Range Dr. James C. Lamb, III Central Power & Light 313 Woodhaven Road P. O. Box 2121 Chapel Hill, North Carolina 27514 Corpus Christi Texas 78403 R. L. Hancock Dr. Emeth A. Luebke Director of Electrical Utilities Atomic Safety & Licensing Comission City of Austin U. S. Nuclear Regulatcry Comission P. O. Box 1088 Washington, D. C. 205t,i .

Austin, Texas 78767 T. H. Muehlenbeck Citizens for Equitable Utilities City of Austin c/o Ms. Peggy Buchorn P. O. Box 1088 Route 1, Box 1684 Austin, Texas 78767 Brazoria, Texas 77422 J. B. Poston Pat Coy Asst. General Manager of Operations Citizens Concerned About Nuclear Power City Public Service Board 5106 Casa Oro P. O. Box 1771 San Antonio, Texas 77422 San Antonio, Texas 78296 A. vonRosenberg Betty Wheeler City Public Service Board Hoffman, Steeg & Wheeler P. O. Box 1771 1008 S. Madison San Antonio, Texas 78296 Amarillo, Texas Brian E. Berwick Bernard M. Bordenick Asst. Attorney for the State of Texas Hearing Attorney P. O. Box 12548 Office of the Executive Legal Director Capitol Station U. S. Nuclear Regulatory Comission Austin, Texas 78711 Washington, D. C. 20555

9 CLADDING SWELLING & RUPTURE MODELS FOR LOCA ANALYSIS A. Evaluation of the potential impact of using fuel rod models pre-

. sented in draft NUREG-0630 on the Loss of Coolant Accident (LOCA) analysis for South Texas Project, Units 1 & 2.

This evaluation is based on the limiting break LOCA analysis identi-fied as follows:

BREAK TYPE DOUBLE ENDED COLD LEG GUILLOTINE BREAK DISCHARGE COEFFICIENT 1.0 WESTINGHOUSE ECCS EVALUATION MODEL VERSION FEBRUARY 1978 CORE PEAKING FACTOR 2.5 HOTRODMAXIMUMTEMPgRATURECALCULATEDFORTHEBURSTREGIONOFTHE CLAD 1891.4 F = PCT B ELEVATION 7.0 Feet.

HOTR0DMAXIMUMTEMPERATgRECALCULATEDFORANON-RUPTUREDREGIONOF THE CLAD 2055.8 F = PCT N ELEVATION 8.75 Feet CLAD STRAIN DURING BLOWDOWN AT THIS ELEVATION 3. 970 Percent MAXIMUM CLAD STRAIN AT THIS ELEVATION 3. 970 Percent Maximum temperature for this non-burst node occurs when the core reflood rate is less than 1.0 inch per second and reflood heat transfer is based on the steam cooling calculation.

AVERAGE HOT ASSEMBLY R0D BURST ELEVATION 7.0 Feet HOT ASSEMBLY BLOCKAGE CALCULATED 47.0 Percent

1. BURST N0DE The maximum potential impact on the ruptured clad node is expressed in terms of the change in the peaking factor limit (FQ) required to maintain a peak clad temperature (PCT) of 2200.00 F and in terms of a change in PCT at a constant FQ (from the Westinghouse letter to the NRC, dated December 7,1979; ref.

NS-TMA-2174). Since the clad-water reaction rate increases significantly at temperatures above 2200.0 F, individual effects (such as APCT due to changes in several fuel rod models) indicated here may not accurately apply over large ranges, but a simultaneous change in FQ which causes the PCT to remain in the l 0

neighborhood of 2200.0 F justifies use of this evaluation procedure.

l l

1

From the December 7,1979 Westinghouse letter to the NRC (ref.

NS-TMA-2174) the following is provided:

t For the Burst Node of the clad:

0 i 0.01 A FQ = ~ 150.0 F BURST N0DE A PCT

- Use of the NRC burst model and the revised Westinghouse burst model could require an FQ reduction of 0.027.

- The maximum estimated imoact of using the NRC strain model is a required FQ reduction of 0.03.

Therefore, the maximum penalty for the Hot Rod burst node is:

0 APCT g = (0.027 + .03) (150.0 F/.01) = 855.0 F 0

Margin to the 2200.0 F limit is:

0 APCT2 = 2200.0 F - PCTB PCT = 1891.4 F B

0 0 APCT2 = 2200.0 F - 1891.4 F = 308.6 F The FQ reduction required to maintain the 2200.0 F clad temperature limit is:

AFQB = (6 PCT 1 - A PCT 2 ) F g .01 A F0) 150.0 F

= ( 855.0 - 308.6) g .01 )

150.0

= 0.0364 (but not less than zero).

2. NON-BURST NODE ,

The maximum temperature calculated for a non-burst section of clad typically occurs at an elevation above the core mid-plane during the core reflood phase of the LOCA transient. The potential impact on that maximum clad temperature of using the NRC fuel rod models can be estimited by examining two aspects of the analyses. The first aspect is the change in pellet-clad gap conductance resulting from a difference in clad strain at the non-burst maximum clad temperature node elevation. Note that clad strain all along the fuel rod stops af ter clad burst occurs and use of a different clad burst model can change the time at which burst is calculated. Three sets of LOCA analysis results were studied to establish an acceptable sensitivity to apply generically in this evaluation. The possible PCT increase 0 resulting from a change in strain (in the Hot Rod) is 20.0 F per percent decrease in strain at the maximum clad temperature locations. Since the clad strain calculated during the reactor

coolant system blowdown phase of the accident is not changed by the use of NRC fuel rod models, the maximum decrease in clad strain that must be considered here is the difference between the " maximum clad strain" and the " clad strain during blowdown" indicated above.

Therefore:

F) (MAX STRAIN - BLOWDOWN STRAIN) ,

APCT3=( *

.01 strain

• ( 20.0)0F (.0397 .0397)

.01

= 0.00F The second aspect of the analysis that can increase PCT is the flow blockage calculated. Since the greatest value of blockage '

indicated by the NRC blockage model is 75 percent, the maximum PCT increase can be estimated by assuming that the current level of blockage in the analysis (indicated above) is raised to 75 percent and then applying an appropriate sensitivity formula as shown in the December 7,1979 Westinghouse letter to the NRC  !

(ref. NS-TMA-2174).

Therefore, 0

APCT4 = 1.25 F (50 - PERgENT CURRENT BLOCKAGE)

+2.36 F (75-50) ,

= 1.25 F (50 - 47.0) + 2.36 F (75-50)

= 62.750F If PCTu occurs when the core reflood rate is greater than 1.0  ;

0 inch par second, APCT4 = 0.0 F. The total potential PCT increase for the non-burst node is then l 0 0 A PCTS = APCT3 + APCT4 = 0.0 F + 62.75 F = 62.75 F 0

Margin to the 2200.0 F limit is 0

APCT6 = 2200.0 F - PCTN F = 2200.0 F - 2055.80F = 144.2 0F The FQ reduction required to maintain this 2200.00 F clad tenperature limit is (from NS-TMA-2174)

AFQN = (A PCT S - APCT6 ) F (.01AFQ )

0 10.0 F APCT AFQN

= -0.08145 (but not less than zero).

The peaking factor reduction required to maintain the 2200.0 F clad temperature limit is therefore the greater of AFQB and A FQN '

or; A FQPENALTY = .0364 B. The effect on LOCA analysis results of using improved analytical and modeling techniques (which are currently approved for use in the Upper Head Injection plant LOCA analyses) in the reactor coolant system blowdown calculation (SATAN computer code) has been quantified via an analysis which has recently been submitted to the NRC for review. Recognizing that review of that analysis is not yet complete and that the benefits associated with those model improvements can change for other plant designs, the NRC has ,

established a credit that is acceptable for this interim period to help offset penalties resulting from application of the NRC fuel rod model s. That credit for two, three and four loop plants is an increase in the LOCA peaking factor limit of 0.12, 0.15 and 0.20, respectively.

C. The peaking factor limit adjustment required to justify plant operation for this interim period is determined as the appropriate LFQ credit identified in section (B) above, minus the AFQ calculated in section ( A) above (but not greater than zero).

FQ ADJUSTMENT = 0.20 - 0.04 = 0.16 I

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