ML19339A556
| ML19339A556 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 10/29/1980 |
| From: | Nichols T SOUTH CAROLINA ELECTRIC & GAS CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0630, RTR-NUREG-630 NUDOCS 8011040268 | |
| Download: ML19339A556 (6) | |
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SOUTH CAROLINA ELECTRIC a GAS COMPANY post omcs een,e4 COLUMBIA, SOUTH CAROUNA 29288 T. C. NicMots,J n.
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. s c..com.
October 29, 1980 no....o i-Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nacicar Regulatory Commission Washington, D. C. 20555
Subject:
Virgil C. F;mmer Nuclear Station Docket No. 50/395 NUREG-0630, Fuel Clad Swelling and Rupture !!odels for LOCA Analysis
Dear Mr. Denton:
In Mr. R. L. Tedesco's letter to South Carolina Electric and Gas (SCE&G) dated 8/26/80, SCE&G was requested to provide supplemental calculations to show conformance to 10CFR50.46 using the fuel rod models presented in NUREG-0630.
In their letters NS-TMA-2147 dated 11/2/79, NS-TMA-2163 dated 11/16/79, NS-TMA-2174 dated 12/7/79 and NS-TMA-2175 dated 12/10/79, Westinghouse has provided the NRC with information on this subject. Presented in the attachment to this letter are the addi-tional calculations for the Virgil C. Summer Nuclear Station requested in Mr. Tedesco's letter.
Westinghouse and SCE&G believe the current Westinghouse models to be conservative and in compliance with Appendix K and the information provided in this letter is sufficient to resolve this item for the Virgil C. Summer Nuclear Station.
If you have any questions, please let us know.
Very truly yours, fa T..C.
Nichols, Jr.
RBC:TCN:rh cc: Page Two
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4 Mr. Harold R. Denton October 29, 1980 Page Two
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cc:
V. C. Summer
.G. H. Fischer 4
l T. C..Nichols, Jr.
E. H. Crews, Jr.
D. A.' Nauman 1
O. S. Bradham O. W..Dixon, Jr.
R. B. Clary 4
-W. A. Williams, Jr.
B. A. Bursey J.~ B. Knotts J. L. Skolds NPCF/Whitaker 3
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EVALUATION OF THE P0TENTIAL IMPACT OF USING FUEL R0D MODELS PRESENTED IN DRAFT NUREG-0630 ON THE LOSS OF COOLANT ACCIDENT (LOCA) ANALYSIS FOR VIRGIL SUMMER,R.A'JT_;(CGE)
-A.
This evaluation i
- sed on the' limiting break LOCA analysis identi-fied as follows:..
r BREAK TYPE:
DOUBLE ENDED COLD LEG GUILLOTINE
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BREAK DISCHARGE COEFFICIENT:
0.6 WESTINGHOUSE ECCS EVALUATION MODEL VERSION:
FEBRUARY 1978
,, CORE PEAKING FACTOR:
2.32 HOT R0D MAXIMUM TEMPERATURE CALCULATED FOR THE BURST REGION OF THE CLAD:
1822.5'F = PCTB ELEVATION:
6.0 Feet.
HOTRODMAXIMUMTEMPERATUREC5LCULATEDFORANON-RUPTUREDREGIONOF THE CLAD:
2145"F - PCTN ELEVATION:
7.5 Feet.
CLAD STRAIN DURING BLOWDOWN AT THIS ELEVATION:
4.136 Percent MAXIMUM CLAD STRAIN AT THIS ELEVATION:
4.756 Percent The maximum temperature for this non-burst node occurs when the core reflood rate is less than 1.0 inch per second and the reflood heat transfer is based on the steam cooling calculation.
_ AVERAGE HOT ASSEMBLY ROD BURST ELEVATION:6.0 Feet.
.H0T ASSEMBLY BLOCKAGE CALCULATED:
42.5 Percent 1.
BURST N0DE The maximum potential impact on the ruptured clad node is expressed in letter NS-TMA-2174 in terms of the change in the peaking factor limit (F ) required to maintain a peak clad temperature (PCT) of Q
J2200*F and in terms of a change in PCT at a constant FQ.
Since the clad-water reaction rate' increases significantly at. temperatures above 2200*F, individual effects 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*F justifies use of this evaluation procedure.
Frcm NS-TMA-2174 for the Burst Node of the clad:
0.01 AFQ is equivalent to - 150*F BURST N0DE APCT I
U2e 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 required FQ reduction ;of 0.03.
. __,__, Therefor ^, the maximum penalty for the _ Hot Ro'd burst node.is:_ _
aPCT1 = (0.027 +.03) (150*F/.01) = 855*F The margin to the 2200*F limit is:
'aPCT2 = 2200*F - PCTB = 377.5'F The FQ reduction required to maintain the 2200*F clad temperature limit isi C
y - APCT ) (.01_aFQ)
AFQB = (APCT 2
150'F
= (855 - 377.5) (.01)
T55
= 0.0318 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 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 increase
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resulting from a change in strain (in the Hot Rod) is +20 *F per percent decrease in strain at the maximum clad temperature loca-tions.
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 at the end of RCS blowdown" indicated above.
Therefore:
APCT3 = (20*F
) (MAX STRAIN - BLOWDOWN STRAIN)
.01 strain
- (20 ) (.04756
.04136) = 12.4*F 0T 4
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i 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 shown in.NS-TMA-2174..-....-
4 Therefore, APCT4 = 1.25'F (50 - PERCENT CURRENT BLOCKAGE)
+ 2.36*F (75-50)
= 1.25 (50 - 42.5) + 2.36 (75-50)
= 68.4*F If PCTN occurs when the core reflood rate is greater thar 1.0 inch per second, APCT4 = 0.
The total potential PCT increase for the non-burst node is then APCTS = APCT3 + APCT4 12.4 + 68.4 = 80.8'F The margin to the 2200*F limit is APCT6 = 2200*F - PCTN = 2200 - 2145 = 55*F The FQ reduction required to maintain this 2200*F clad temperature limit is (from NS-TMA-2174) 5 - APCT ) (.01aFQ
)
AFQN = (APCT 6
T07 APCT AFQN = 0.0258
".he peaking factor reduction required to maintain the 2200*f clad
- ,'- temperature-limit.is..therefore the greater of AFQB and AFQN.
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or; a FQPENALTY = 0.0318 B.
The effect on LOCA analysis results of using improved analytical'and
'modeling techniques (which are currently approved for use in the i
Upper Head Injection plant LOCA analyses) in the reactor coolant system blowdown calculation (SATAN computer code) has been quanti-fied via an analysis which has recently been submitted to the NRC for review.
Recognizing tr.at review of that an.alysis is not yet complete and that the benefits associatec wito those model improve-ments can change for other plant designs, the KRC has established a credit that is acceptable for this interim period to help offset penalties resulting from application of the NRC fuel rod models.
4 That credit for three loop plants is an increase in the LOCA peaking factor limit of 0.15.
~
C.
The peaki.ng factor limit adjustment' required to justify plant opera-tion for this interim period is determined as the appropriate AFQ credit identified in section (B) above, minus the AFQPENAi.TY cal-culated in'section'(A) above but.not greater than zero).
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FQ ADJUSTMENTe7 0'.15T20'.0318 - 0.~0:1:_~l: ~."I~~
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