ML20003C431
| ML20003C431 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 02/27/1981 |
| From: | Goldberg J HOUSTON LIGHTING & POWER CO. |
| To: | Eisenhut D NRC |
| References | |
| ST-HL-AE-623, NUDOCS 8103050658 | |
| Download: ML20003C431 (6) | |
Text
.
The Light COE%y Houston Lighting & Power P.O. Box 1700 Houston. Texas 77001 (713)228-9211
_ _. _ _....., ~... _ _.. _ _ -... - ~. - - -
February 27, 1981 ST-HL-AE-623 SFN: C-0100 V-0100
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SQh Vx Mr. Darrell G. Eisenhut CY
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Division of Project Management
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' [_g hu: lear Regulatory Comission
,7, Washington, D. C.
20555 l
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Dear Mr. Eisenhut:
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South Texas Project dtU
.c y 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 infonnation 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 infon11a-
+
tion which utilizes the materials models of draft NUREG-0630.
In response to the above mentioned request, attached is the evaluation 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. TN South Texas Project (STP) FSAR currently reflects the nine (9) grid fuel design and the STP-FSAR will be a;.1 ended 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.
Ver., truly yours, O[
{U.'lh]fe J. H. Goldberg I[
Vice President Nuclear Engineering & Construction MEP/ par Attachment 810 so so fos 8-A
Houston Lighting & Power Company February 27, 1981 cc:
J. H. Goldberg ST-HL-AE-623 D. G. Barker SFN: C-0100 V-0100 Howard Pyle 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 & Eaforcement Nuclear Regulatory Comission Washington, D. C.
20555 M. L. Borchelt Charles Bechoefer, Esquire Executive Vice President Chaiman, 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 Cer4 tral 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 Commission City of Austin U. S. Nuclear Regulatory Comission P. O. Box 1088 Washington, D. C.
20555 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 1584 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 78295 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
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.
~his evaluation is baser. on the limiting break LOCA analysis identf-ficd 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 = FCTB 5
ELEVATION 7.0 Feet.
HOTR0DMAXIMUMTEMPERATgRECALCULATEDFORANON-RUPTUREDREGIONOF THE CLAD 2055.8 F = PCT f
N ELEVATION 8.75 Feet r
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 i
HOT ASSEMBLY BLOCKAGE CALCULATED 47.0 Percent 1.
BURST NODE i
The maximum potential impact on the ruptured clad node is expressed in terms of the change in the peaking factor limit (FQ) required to ma' tain a peak clad temperature (PCT) of 2200.0 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 ? ? action rate increases significantly at temperatures above 2;.30.0 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 neighborhood of 2200.0 F justifies use of this evaluation procedure.
4 From the December 7,1979 Westinghouse letter to the NRC (ref.
NS-TMA-2174) the following is provided:
For the Burst Node of the clad:
0 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 impact of using the NRC strain model is a requirdd FQ reduction of 0.03.
Therefore, the maximum penalty for the Hot Rud burst node is:
0 0
APCT
= (0.027 +.03) (150.0 F/.01) = 855.0 F y
0 Margin to the 2200.0 F limit is:
0 0 F APCT2 = 2200.0 F - PCTg PCT
= 18 1.4%
B 0
0 0
APCT2 = 2200.0 ' - 1891.4 F = 308.6 F 0
The FQ reduction required to maintain the 2200.0 F clad temperature limit is:
AFQB = (A PCTy - 6 PCT ) F g.01 A FQ) 2 0
150.0 F
= ( 855.0 - 308.6) g.01 )
150.0
= 0.0364 (but not less than zero).
2.
NON-BURST N0DE 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 estimated 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 after 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 increase0 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 curing the reactor 1
J
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:
APCT3=(
F)
( m S M IN - R0WDOWN S M IN)
.01 strain
= ( 20.0) F (.0397
.0397)
.01 0
= 0.0 F 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 i
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, i
0 APCT4=1.25F(50-PERgENTCURRENTBLOCKAGE)
+2.36 F (75-50) 0 0
= 1.25 F (50 - 47.0) + 2.36 F (75-50) 0
= 62.75 F If PCT occurs when the core reflood rate is greater than 1.0 N
g inch per second, APCT4 = 0.0 F.
The total potential PCT increase for the non-burst node is then 0
APCT5 = APCT3 + APCT4 = 0.0 F + 62.75 F = 62.75 F 0
Margin to the 2200.0 F limit is 0
U F = 2200.0 F - 2055.8 F = 144.2 F APCT6 = 2200.0 F - PCTN 0
The FQ reduction required to maintain this 2200.0 F clad temperature limit is (from NS-TMA-2174)
AFQ = (APCT5 - APCT ) F (.01aF0
)
6 N
0 10.0 F APCT AFQ = -0.08145 (but not less than zero).
N
?
1
t-The peaking factor reduction required to maintain the 2200.0 F 0
clad temperature limit is therefore the greater of AFQ and A l
B FQ '
N 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 a. 'ociated 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 NPC 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 LFQ calculated in section ( A) above (but not greater than zero).
FQ ADJUSTMENT = 0.20 - 0.04 = 0.16 l
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