ML19327A295

ECCS Performance Analysis W/Rate Dependent Burst Model.

ML19327A295
Person / Time
Site:	Sequoyah
Issue date:	07/31/1980
From:	WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML19327A294	List: ML19327A293 ML19327A295
References
NUDOCS 8008050320
Download: ML19327A295 (10)

Text

.

ENCLOSURE SEQUOYAH NUCLEAR PLANT UNIT 1 ECCS Performance Analysis with Rate Dependent Burst Model l

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8008050g)0 -

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The fluclear Regulatory Cc=nission (NRC) issued a letter dated tiovember 9,1979 to operators of light water reactors regarding fuc1 rod recels used in loss of Ccolant Accident (LOCA) ECCS evaluation medels. That letter describes a meeting called by the tac en ilcvember 1, ~1979 to present draf t repcet ::UREG 0620, " Cladding Swelling and Ruoture Models for LOCA Analysis." At the-meeting, representatives of tiSSS vencors and fuel suppliers wre asked to show how plants licensed using their LOCA/ECCS evaluatien mocel continued to ccnform to 10 CFR Part 50-46 in view of the new fuel red models presented in draf t ;;UREG 0520. . ,

Westinghcuse representatives presented informaticn on the fuel red models used in analyses for plants licensed with the Westinghcusa ECCS evaluation model and discussed the potential impact _cf fuc.1 red r.cdel changes on results of those analyses. That information was fermally documented in letter !!S-TG-2147, dated liovember 2,1979, and force'd the 4

basis for the Westinghcuse conclusion that the information presented in draf t I;UREG 0520 did not constitute a saf ety problem f or Westinchcuse

~

plants and that all plants confermed with 1:RC reculations. In the flovember 9,1979 letter, the 22C requested that 5per:: rs of licht water

-react' ors provide, within sixty (60) days, information which wili enable

'the staff to determine, in light of tne fuel red model concerns, whether or not further action is necessary.

This letter provides information en the LOCA analysis cf your plant required to respcnd to that request. Note, ho.ever, that a significant amount of discussion and inf ormation exchange between Westinghouse and the !!RC has transpired since the :levember 2 letter (:S-D%-21*7) was prepared and the basis for demcastrating ccepliance with 10 CFR part 50 ,

has been modified." The following is an outline of the significant events that occurred since ilcycmber 2 and is providad to update you on this situation. -

As a result of compiling information for letter tis-T:%-2147 Westinghouse recogniz!d a potential discrepancy in the calculation of .

fuel red burst fer cases having clad heatup ratas (;rior to rupture) significantly lower than 25 degrees F per second. This issue was "

, reported to. the NRC staff, by telephene, on :lovember 9,1979, and although independent of the !2C fuel red model concern, the combined effect of this issue cnd the effect of the fiRC fuel red models had to Be

studi ed. Details of the work done on this. issue were presented to the

- - ~

NRC on !cvember 13, 1969 and documented in letter ::S-T!%-2163 dated November 16,'19797 That work included development of a procedure to

. determine the clad haatup rate prict to burst and a reevaluation of *

cperating 'c!cstinghcuse plants with censideration of a mcdified -

Westinghouse fuel red burst model. As part of this. reevaluation, the Westinghouse position en :tUREG-CrS30 vas rev_iewed and it was still con- '

cluded that the information presentec in draf t tiUREG-0530 dio not ccn-stitute a safety problem for plants licensed with the Westinghouse ECCS evaluation medel.

On December 6,1979, :RC and Westinghouse personnel discussed the infor-l mation thus far prtsanted.. At the conclusion of that discussion, the l NRC staff requested : estinghcuse to revide further detail on the coten- -

tial impact of mocificaticns to acen of the fuel red mcdels usec in.the j LOCA analysis anc to outline analytical reedel inprov tments in other ,

i parts of the analvsn ane rh.s enn. -i ,3 ',c na u - < n u . m a,,e

improvemen ts.- This additional information was complied frem varicus LOCA -

analysis results and documented in letter NS-T!%-2174 dated December 7, ~

1979.-

---

~ jnother'meetingwasheldinBethesdaenDecember 20, 1979 where NRC an'd Westinghouse personnel estab_lished: 1) The currently accepted procecure for assessing :he potential impact cn LOCA analysis results of using the

-- 4' fuel rod models presented in draf t i:UREG-0620 and 2) Acceptable benefits m ;.. resulting frem enalytical model improvements that would justify centinued plant operatien for the interim until differences between the fuel rod .

medcls of concern are resolved. __ .-

The information following on pages 3 - 6 is expected to satisfy the ::RC request for information en your plant which will enable the f:RC to deter-mine whether or not further action is necessary.

Part of the Westinghcuse effort provided to assist in the resolution of these1LOCA fuel rod medel differences is documented in letter NS-T:4A-2175, date'd December 10, 1979, which contains Westinghcuse coments en draf t NUREG-0530. As stated in that letter, Westingneuse believes the current Westinghouse medels to be conservative and to be in ccmpi;snce with Appendix K. -

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A.

Evaliiation of the potentiai impact of using fuel rod models presented in draft NUREG-0630 en the loss of Coolant Accident (LOCA) analysis for

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Sequoyah.

This evaluatien is based on the limiting break LOCA analysis identified as fol!cws: .-.

' BREAK TYPt DOUBLE ENDE0 COLD LEG GUILLOTIriE_ .

0.6 IMP MIXING BREAK OISCHARGE COEFFICIENT -

FEB '7P._ - _ _

4 WESTINGHOUSE ECCS EVALUATION HCCEL VERSION CORE FEAKING FACTOR 2.25

--HOTRODPAXIMUMTEMPERATgRECALCULATEDFORTHEBURSTREGIONOFTHE -- -

CLAD - 1705 F = PCTg'

• 'h Feet - ~.

ELEVATION - 6.5 HOT RCD MAXIMUM TEMPE? URE CALCULATED FOR A NON-RUPTURED REGION OF THE CLAD - 2143 F = PCT N Feet ELEVATION - _ 7.5 ,

10 Percent CLAO' STRAIN DURING BLOWDC'a'N AT10THIS ELEVATION Percent

. MAXIMUM CLAD STRAIN AT THIS ELEVATION Maxicum 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.

l Feet AVERAGE HOT ASSEF3LY ROD BURST ELEVATION - 7.25 '

34.9 Percent .

' HOT ASSEMhtY BLOCKAGE CALCULATED - . l

1. BURST NODE.

~

l' The ' maximum potential impact on the ruptured clad code is expressed in letter NS-TMA-2174 in terms of the change in the peaking factor ~

limit' (FQ) required to maintain a peak clad temperature (PCT) Sinceof 22000F and in terms of a change in PCT at a constant FQ.

the clad-water reaction rate increases significantly at temperatures I above 2200.0F, individual effects (such as 6 PCT due to changes in l several fuel red 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 2200.0F justifies use of this evaluation procedure. .

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, From N5.TMA-2174:

For the Burst tiode of tne clad:,

' O 01 aFQ - s 150 F BURST N00E APCl Use of the NRC burst rodel and the revised Westin house burst model could require an FQ reduction of 0.027 The maximte estimated impact 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 aPCTg = (0.027 + .03) (l50 F/.01) = 855 F -- --

Hargin to the 2200 F limit is:

APCT 2

= 2200.0F - U0S PCT 3

= G5 9 The FQ reduction required to maintain the 2200 0F clad temperature

... limi t is : ,,

aFQg=(APCT;-APCT) 2 I' )

150*F

= (855 - g ) ( )

=

.024 (but not less than zero).

2. NON-BUP.5T 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 esximum 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 red stops after clad burst occurs ar.d use of a different clad burst model can change the time at'which burst is calculated.

The effect of the change in pellet-clad gap conductance in the non- '-

burst node was determined by performing a sensitivity study using various rate deoendent burst curves. The results of these studies are presented in Table I. It is seen that the most representative case, asdefinedbytherelationshipbetweentheactualcladgeatuprate and the heatup rate used in the burst model, is the 8.4 F/sec ramp rate. The increase in calculated peak clad temperature, APCTs, for' this case was 19eF. It is also noted that hot rod burst occurs 6.9 seconds earlier and the blockage increases by 9.7% over the base '

case results.

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The sensitivity study included the effect"of bicekage as well as the effect of the change in gap conductance on the peak clad tercerature.

Thus in The value of APCT 3 includes an ir. creased blockage penalty; the evaluation of the increase in PCT, APCT4 , due to bicekage, the 45% blockage value from the sensitivity rather than the 3 .9% valve from the limiting break LOCA analysis identified at the beginning of section A will be used in order to prevent a dcubie acccunting of the 9.7% blockage increase noted in the sensitivity :cudy.

study A comparison of the ramp' dependent burst curves used in thi:

with the NUP.EG-0630 is shown in Figure A-1. It can be seen that the burst temperatures predicted by the K rate dependent models predict

" higher burst temperatures than the ORNL correlat-ions.~ Detailed discussions of these differences are provided in the Decerter 10 letter (HS-TMA-2175), including an explanation of the censervatism of the rodel and demonstration of compliance With Appendix K.

The second aspect of the analysis that can increase PCT is the flow

- ' -blockage calculated. Since the greatest value of bicekage indicated by

- -'- the NRC blockage ocdel is 75 percent, the maximum PCT increase can be estimated by assuming that the current level of bicckage in the analysis (see above discussion) is raised to 75 percent and tnen applying an apprcpriate sensitivity forcula shown in NS-TMA-2174.

Therefore, 0

APCT,g = 1.25 F (50 - PERCENT CURRENT BLOCKAGE) '

+ 2.36 F (75-50)

= 1.25 (50 - 45) + 2.36 (75-50)

= 65.25 F If PCTg occurs when the core reficed rate is greater than 1.0 inch per seccnd AFCT = 0. The total potential PCT increase for the non-burstnodeist$en ,

- + hCT; = 19.0 + 65.25 = 84.25 F aCT3 = APCT3 Margin to the 2200 F limit ~ is . ,

APCT 6

= 2200 F - PCTH = 2200 - 2145 = 57 F 0

The FQ reduction required to maintain this 2200 F clad temperature limit is (from NS-TMA-2174) aFO N

=

(a CT5~# 6) ) = (8 25 - 57)(h = .027 10 F APCT

.027 but not less than zero. -

=  ;

AFQN 0

- The peaking factor reduction required to maintain the 2200 F clad temperature limit is therefore the greater of FQ3 and aF03 ,

or; A FQ = .027 PENAL .

S. The NF.C has recently approved the removal of the 65 F uncertainty -

cn the hot red fuel pellet temperature for ECCS analysis. The effect of removing this uncertainty on the calculated PCT has been determined based on previously establ'ished sensitivities performed to quantify this effect (WCAP-9180). From these, it is estimated that this reduction in applied model uncertainty wuld result in a decrease in calculated PCT of 15CF for UHI plants.

- Applying the same sensitivity used in calculating aFQN '

aF0 CREDIT 15 F C ),. e .01.57-10 F AFCT ,,

C. The peaking factor limit adjustment required to justify plant operation for this interim period is deter =ined as the apprcpriate

- aFQ calculated credit in identified section (A) in section above (but (B) above, not greater minus the:erc than  !.FQ,ig77

~

= .012 iQ ADJUSTPENT = .015. .027 D. The revised peaking factor is then FQ FSAP minus the FQ adjust.ent, or:

FQ = 2.25' ~

.012 = 2.237 -

4 E. The results of the "18 case' analyses are shown in figure 2.

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