Text

License Change Application 60 to Amend License NPF-1. Provides Revised ECCS Evaluation & Replaces License Change Application 51
	ML19309H465
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	05/07/1980
From:	Goodwin C; PORTLAND GENERAL ELECTRIC CO.
To:	;
Shared Package
ML19309H463	List: ML19309H462; ML19309H465;
References
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	Download: ML19309H465 (75)
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l PORTLAND GENERAL ELECTRIC C0!!PANY EUGENE WATER & ELECTRIC BOARD AND PACIFIC POWER & LIGHT C0!!PANY TROJAN NUCLEAR PLANT Operating License NPF-1 Docket 50-344 License Change Application 60 a

This License Change Application provides a revised Emergency Core Cooling System performance evaluation for the Trojan plant using the latest j

NRC-approved Westinghouse Evaluation !!odel (February 1978 version).

J f

PORTLAND GENERAL ELECTRIC C0t!PANY l

/

By f

C. Goodwin, Jr.[/

Assistant Vice President Thermal Plant Operation and Maistenance Subscribed and sworn to before me this 7th day of !!ay 1980.

DJ u

Notary Public of Q(egon

[u,.s,,mf f. /((J

!!y Commission Expires:

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I.

l LCA 60 Attachment A j

Page 1 of 3 LICENSE CRANCE APPLICATION 60 i

l DESCRIPTION OF CHANGE 1

To address the latest fuel rod model concerns and at the same time recover the margin that has been lost in connection with previous Emergency Core Cooling System (ECCS) Evaluation Model changes, and to accommodate the possibility of plugging of as many as all Row I and Row 2 steam generator tubes, a new ECCS analysis has been performed for Trojan using the latest j

NRC-approved Westinghouse Evaluation Model (February 1978 version).

l This revised ECCS performance evaluation consists of two parts. The first provides a tabular summary of relevant input assumptions in the large LOCA analysis and calculated output parameters.

In addition, plots of key parameters during the course of the LOCA event are provided.

The 4

second part is a quantitative discussion of the impact of incorporating l

NRC f uel rod models into the Westinghouse Evaluation Model.

This dis-cussion responds to the concern about fuel rod modeling and is consistent with current NRC policy regarding ECCS analyses pending final agreement on fuel rod modeling questions.

In summary, the revised ECCS evaluation is consistent with unrestricted operation of Trojan with up to 6 percent uniform steam generator tube t

plugging, with a maximum total peaking f actor of 2.32.

Under these conditions, the calculated PCT during the worst-case LOCA is 1970*F.

The impact of incorporating NRC fuel rod models is demonstrated to be more I

than compensated by inherent conservatisms in the Westinghouse Evaluation Model.

Therefore, no reduction in peaking factor or other operational restrictions are necessary to address the fuel rod modeling concerns.

REASON FOR CHANGE f

Demonstration that the Plant's ECCS meet certain performance standards is a requirement of 10 CFR 50.46.

This requirement has previously been met for Trojan by submitting ECCS performance analyses using NRC-approved d

Westinghouse ECCS Evaluation Models.

Since the original licensing of Trojan, a number of shortcomings have been identified by Westinghouse and the NRC in the Westinghouse ECCS Evaluation Models; this has resulted in the need to reanalyze ECCS performance using improved models with the errors corrected. For example, a Trojan reanalysis was performed to accommodate the fact that the water temperature in the reactor vessel r

upper head region was initially thought to be intermediate between the hot-and cold-leg temperatures, but subsequent experiments showed it to 1

be closer to the hot-leg temperature. Also, a reanalysis was done to correct an inconsistency in the handling of the heat input from the zirconium-water reaction.

The most recent ECCS Evaluation Model concern was identified in 1979 involving acceptability of Westinghouse fuel rod burst and strain modeling in light of recent experimental results.

The overall result of these ECCS reanalyses has been to decrease the margin between the calculated Trojan ECCS performance and the limits set forth in 10 CFR 50.46.

The current situation for Trojan, even before i

LCA 60 Page 2 of 3 considering the fuel rod modeling concerns, is that it is no longer possible to meet the peak cladding temperature (PCT) limit of 2200*F with the original maximum peaking f actor (Fq) of 2.32.

A Technical Specifi-cation change request is pending at the NRC (LCA 51) to reduce the allowable maximum Fq to 2.25 in order to meet the 2200*F PCT limit, and Fq is presently administrative 1y limited to 2.23.

A further reduction in Fq could potentially impact on operational flexibility by prohibit-

)

ing certain load-following maneuvers.

Furthermore, the current Trojan ECCS performance evaluation is based on the October 1975 version of the Westinghouse Evaluation Model which has been superseded by the February 1978 version.

LCA 51 is to be withdrawn since the evaluation reported' in the present submittal demonstrates that an Fq reduction is not required.

I Steam generator tube plugging also has a detrimental af fect on ECCS pe rfo rmance.

Plugging leiels cf up to several percent of the steam generator tubes are generally acceptable, however an analysis of ECCS performance must be made to demonstrate that the plugging does not in fact significantly degrade ECCS performance.

Based on the Trojan steam generator inspections in late 1979, there is a possibility that addi-tional Row I steam generator tubes may have to be plugged.

Therefore, the ECCS evaluation reported in this LCA accounts for an appropriate level of plugged steam generator tubes.

SAFETY / ENVIRONMENTAL EVALUATION This revised ECCS performance evaluation has been made using an NRC-approved evaluation model which satisfies the' requirements of 10 CFR 50.46 and Appendix K.

The currently applicable small-break LOCA analyses for Trojan are unaffected by the considerations that have led to having the 1

design basis LOCA analyses performed.

Therefore, it is appropriate tha t the current ECCS performance evaluation consists only of a spectrum of worst-case large-break LOCAs.

The methods for accounting for steam generator tube p!' dging are an integral part of the Westinghouse evalu-ation model and have been approved by the NRC.

The discussion of the impact of incorporating NRC fuel rod models is consistent with current NRC policy and adequately responds to the concern.

The revised ECCS analyses were made with an evaluation model in which certain calculational errors and inconsistencies have been corrected.

No change in actual Plant operating conditions is intended; it is only intended that the analysis of the ECCS during Plant operations be correct and appropriately conservative.

The. administrative ' reduction from the current Technical Specification peaking factor limit of 2.32 to a value of 2.23 was made to assure that pending completion of revised analyses using a corrected evaluation model, there would be no chance that the 10 CFR 50.46 limits could be exceeded during a LOCA.

The appropriate calculations have now been com-pleted and have demonstrated that considerable margin exists between the l

regulatory limits and the calculated ECCS performance.

On this basis, relaxation of the administrative Fq reduction is justified and opera-tion up to a peaking factor limit of 2.32 (Technical Specificatian 3.2.2) is acceptable..

+

4 LCA 60 Attachment A Page 3 of 3 SCHEDULE CONSIDERATIONS The ECCS evaluation presented with this LCA provides the basis for relax-ing the current administrative limit on total peaking f actor (Fq) of 2.23 and reverting to the Technical Specification limit of 2.32.

This change in Plant operation will be made upon startup of Cycle 3, which is currently scheduled for July 1, 1980.

CJP/4sa9311

LCA 60 Attachment B 66_Pages FIAT I - Revised Trojan Large Br;ak LOCA Analysis f

4 a

6 O

e The loss of Coolant Accident (LOCA) has been reanalyzed for TA OJAN veri / ACR)

The follcwing information amends the Safety Analysis Report section on Major Reactor Coolant System Pipe Ruptures.

The results are consistent with acceptance criteria provided in reference 1.

The description of the various aspects of the LOCA analysis is given in E3 WCAP-8339 The individual ccmputer ccdes which ccmprise the Westinghouse Emergency Core Cooling System (ECCS) evaluation model are

~0 described in detail in separate reports along with code modifi-cations specified in references 7, 9 and 10.

The analysis presented here was performed with the February 1978 version of the evaluation model which includes modificatiens delineated in references 11, 12, 13 and 14.

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4 Results The analysis of the loss of coolant accident is performed at 102 percent of the licensed core power rating.

The peak linear power and total core power used in the analysis are given in Table 2.

Since there is margin between the value of peak linear power density used in this analysis and the value of the peak linear power density expected during plant oper,a-tion, the peak clad temperature calculated in this analysis is greater than the maximum clad temperature expected to exist.

Table 1 presents the occurence time for various events throughout the accident transient.

Table 2 presents selected input values and results from the hot fuel rod thermal transient calculation. For these results, the hot spot is defined as the location of maximum peak clad temperatures.

That loca-tion is specified in Table 2 for each break analyzed.

The location is indicated in feet which presents elevation above the bottom of the active fuel stack.

Table 3 presents a summary of the various containment systems parameters ano structural parameters which were used as input to the C0CO computer code used in this analysis.

Tables 4 and 5 pres-reflood mass and energy releases to the contain-ment, and the broken

'op accumulator mass and energy release to the containment, respecti ely.

4

The results of several sensitivity studies are reported These results are for conditions which are not limiting in nature and hence are reported on a generic basis.

Figures 1 through 17 present the transients for the principle parameters for the break sizes analyzed.

The following items are noted:

Figures 1A - 3C:

Quality, mass velocity and clad heat transfer coeffi-cient for the h?,tspot and burst locations Figures 4A - 6C:

Core pressure, break flow, and core pressure drop.

The break flow is the sum of the flowrates from both ends of the guillotine break.

The core pressure drop is taken as the pressure just before the core inlet to the pressure just beyond the core outlet Figures 7A - 9C:

Clad temperatur-2, fluid temperature and core flow.

The clad and fluid temperatures are for the hot spot and burst locations Figures 10A - 11C: Downcomer and core water level during reflood, and flooding rate e

Figures 12A - 13C: Emergency core cooling system flowrates, for both accumulator and pumped safety injection e

4 Figures 14A - 15C: Containment pressure and core power transients I

r Figures 16, 17:

Break energy release during blowdown and the contain-ment wall condensing heat transfer coefficient for i

the worst break l

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l t

i L

W l

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1 l

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4 i

Conclusions - Thermal Analysis For breaks up to and including the double ended.severince of a reactor coolant pipe, the Emergency Core Cooling System will meet the Acceptance Criteria as presented in 10CFR50.46 El3 That is:

0 1.

The calculated peak clad temperature does not exceed 2200 F based on a total core peaking f actor of J2.3fL 2.

The amount of fuel element cladding that reacts chemically with water or stesn does not exceed.1 percent of the total amount of Zircalloy in the reactor.

3.

The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling. The cladding oxidation limits of 17% are not exceeded during or af ter quenching.

4.

The core temperature is reduced and decay heat is renoved for an extended period of time, as required by the long-lived radioactivity renaining in the core.

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References for Section 15.4.1

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1.

" Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors",10CFR50.46 and Appendix K of 10CFR50.46. Federal Register, Volume 39, Number 3, January 4,1974.

2.

Bordelon, F. M., Massie, H. W., And Zordan, T.

A., " Westinghouse ECCS Evaluation Model-Summary," WCAP-8339, July 1974.

3.

Bordelon, F. M., et al., " SATAN-VI Program:

Comprehensive Space-Time Dependent Analysis of Lass-of-Coolant," WCAP-8302 (Pro-prietary Version), WCAP-8306 (Non-Proprietary Version)*, June 1974 4.

Bordelon, F. M., Et al., "LOCTA-IV Program:

Loss-of-Coolant Tran -

sient Analysis," WCAP-8301 (Proprietary Version), WCAP-8305 (Non-Proprietary Version), June 1974.

t 5.

Kelly, R. D., et al., " Calculational Model for Core Reflooding af ter a Loss-of-Coolant Accident (WREFLOOD Code)." WCAP-8170 (Proprietary Version), WCAP-8171 (Non-Proprietary Version), June 1974.

6.

Bordelon, F. M., and Murphy E. T., " Containment Pressure Analysis Code (C0CO)," WCAP-8327 (Proprietary Version), WCAP-8326 (Non-Pro-prietary Version), June 1974.

7.

Bordelon F. M., et al., "The Westinghouse ECCS Evaluation Model:

Supplementary Information," WCAP-8471 (Proprietary Version), WCAP-8472 (Non-Proprietary Version), January 1975.

8.

Salvatori, R., "Westingrouse ECCS - Plant Sensitivity Studies,"

WCAP-8340 (Proprietary Version), WCAP-8356 (Non-Proprietary Ver-sion), July 1974.

9. " Westinghouse ECCS Evaluation Model, October,1975 Versions," WCAP-8622 (Proprietary Version), WCAP-8623 (Non-Proprietary Version),

November, 1975.

10. Letter from C. Eicheldinger of Westinghouse Electric Corporation to

0. B. Vassalo of the Nuclear Regulatory Commission, letter number NS-CE-924, January 23, 1976.

11. Kelly, R. O., Thompson, C. M., et. al., " Westinghouse Energency Core Cooling System Evaluation Model for Analyzing Large LOCA's During Operation With One Loop Out of Service for Plants Without Loop Iso-lation Valvea," WCAP-9166, February,1978.

12. Eicheldinger C., " West'nghouse ECCS Evaluation Model, February 1978 Version," WCAP-9220-P-A (Proprietary Version), WCAP-9221-A (Non-Pro-prietary Version), February, 1978.

13. Letter frem T. M. Anderson of Westinghouse Electric Corporation to John Stolz of the Nuclear Regulatory Commission, letter numoer NS-TMA-1981, Nov. 1, 1978.

14. Letter from T. M. Anderson of Westinghouse Electric Corporation to John Stolz of the Nuclear Regulatory Commission, letter number NS-TMA-2014, Dec. 11,'1978.

O

1 TABLE 1 LARGE BREAK - TIME SEQUENCE OF EVENTS EVENT OCCURENCE TIME (SECONDS) j l

DECLG, C = C. </

DECLG, C =06 DECLG, CD" D

D Accident Initiation 0.0 0.0 0.0 Reactor Trip Signal

/. '/</

)., 7,2 1,7/

Safety Injection Signal M7, I C[

.M i, 3a

- U.9 7 j, j Start Accumulator Injection 70 b

/.C 9

13. Y End.of ECC Bypass

'/,)? /

2 7, /O'/

'J e/, / //

End of Blowdown

. f.ul4

.M7. 'J5'

)b,6 Bottom of Core Recovery

@h $l e/O.W7 3?,3'/'/

Accumulators Empty

'U. 21-1 l SQ.6 F/

3'd.f/A' Start Pumped ECC Injection lAr. C/

/7 O'

9 zg../

jf*77

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TABLE 2 LARGE BREAK - ANALYSIS INPUT AND RESULTS Quantitles in the calculations:

Licensed core power rating 102% of 3#///

MWt Total core peaking f actor

'/., 3 2 Peak linear power 102% of l,7, (c.7 kw/ft Accumulator water volume

'/d b cubic feet per tank Accumulator pressure tr 6 6 PSIA Number of Safety Injection Pumps Operating 3

Steam Generator Tube Plugging Level l9 percent (uniform)

Fuel Parameters -

Cycle Region DECLG, C = D, la DECLG, CD" Results DECLG, C

=

~'

D D

Peak clad temperature (O )

157J 6 9

/767.78 19 0f' fr F

Location (feet)

7. #

6.0

7. C Maximum local clad / water reaction (%)

6. 7 %

T,.?_7_

f. 6 6 Location (feet)

_ _ Z,.(

_., _(,,,0 7, C Total core clad / water reaction (%)

2 (. o T

4. O -il

4. A J' Ilot rod burst time (seconds)

,28. d 7 f,, f Location (feet)

(,f

/., O

TABLE 3 CONTAINMENT DATA (nRY CONTAINMENT)

Net Free Volume

),1 (, 5 x so "

Initial Conditions l4.9 MIR Pressure Tcacaratare 3, )

';~

R:4ST Temperat'are 37..

, r-Service Water Temperature 40 f Otiside Temperaturt J

Spray System Number of Pumps Operating 7

Rurout Flow Rate L000 G e n^

Actuation Time 40 Safeguards' Fan Coolers Number of Fan Coolers Operating 2

Fastest Post Accident Initiation of Fan Coolers

y.. _ _ _.

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TABLE 3 (Cant'd)

STRUCTURAL HEAT SINK DATA Thickness (in)

Material Area, ft2 40

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PAINTED STRUCTURAL HEAT SINK DATA Structural Heat Sink Structural Heat Sink Paint Thickness Surface Area (Ft2)

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TABLE 4 REFLOOD MASS AND ENERGY RELEASE Gn]

Time (Sec)

Mass Flow (Lb/Sec)

Eneray Flow (Cn/Sec) 41,6

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TABLE 5 BROKEN LOOP AC,jMULATOR MASS AND ENERGY RELEASE Time (Sec)

Mass Flog (th/Sec)

Energy Flow (Btu /Sec) 0.000 3152.322 182834.657 1.010 2909.912 168774.898 H 4=

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...o 28.711 1250.772 72544.e00 28.912 1247.129 72333.506 29.111 29.310

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s PART II - Fuel Rod Modeling Concerns A.

Evsluation of the combined impact of (1) using NRC fuel rod models presented in draft NUREG-0630, and (2) incorporating a heatup-rate dependent clad burst model on the Loss-of-Coolant Accident (LOCA) analysis for T/Sf4V This evaluation is based on the limiting break LOCA analysis identi-fied as follows:

BREAK TYPE DOUBLE ENDED COLD LEG GUILLOTINE BREAK DISCHARGE COEFFICIENT D6 WESTINGHOUSE ECCS EVALUATION MODEL VERSION /9dd/A<f /878 CORE PEAKING FACTOR 7 J2 HOT ROD MAXIMUM TEMPERATURE CALCULATED FOR THE BURST REGION OF THE CLAD

/978

'F = PCTB ELEVATION 4.0 Ft.

HOT ROD MAXIMUM TEMPERATURE CALCULATED FOR A NON-RUPTURED REGION OF THE CLAD

/959 Ft.

ELEVATION 7,26 Ft.

CLAD STRAIN DURING BLOWDOWN AT THIS ELEVATION O

Percent MAXIMUM CLAD STRAIN AT THIS ELEVATION J,4 J Percent Maximum temperature for this nonburst node occurs when the core reflood rate is greater than 1.0 in./sec and reflood heat transfer is based on the FLECHT calculation.

AVERAGE HOT ASSEMBLY R0D BURST ELEVATION

$. 8 Ft.

HOT ASSEMBLY BLOCKAGE CALCULATED M9 Percent 1.

BURST NODE 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 (Fq) required to maintain a peak clad temperature (PCT) of 2200*F and in terms of change in PCT at a constant Fq.

Since the clad water reaction rate increases significantly at temperatures above 2200*F, indi-vidual 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 FO vhich causes the PCT to remain in the neighborhood of 27Co*F justifies use of this evaluation procedure.

~

o From NS-TMA-2174:

For the Burst Node of the clad:

0.01 AFq + < 150*F BURST NODE APCT Using a Westinghouse burst model revised to factor in heatup rate-dependence could require an Fq reduction of 0.012, and use of the NRC burst model could require a further Fq reduction of 0.015 for a total reduction of 0.027 The maxiumum 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:

APCT1 = (0.027 +.03) (150*F/.01) = 855*F Margin to the 2200*F limit is:

APCT2 = 2200*F - PCTB= 2178 F The Fq reduction is required to maintain the 2200*F clad temperature limit is:

.01 AFq AFQB = (aPCT1 - APCT ) ( 150*F 2

= ( 8ff - 2Jc:7 )(

)

= 0.092 ( 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 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 resulting from a change in strain (in the Hot Rod) is +20*F

o j

5 l

l per percent decrease in strain at the maximum clad tempera-ture 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 at the end of RCS blowdown" indicated above.

Therefore:

APCT3 " ( 01

~

t ain

=(

)(adDJ - o

)

,2,

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 maxiumum 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 appro-priate sensitivity formula shown in NS-TMA-2174.

Therefore:

APCT4 = 1.25*F (50 - PERCENT CURRENT BLOCKAGE)

+ 2.36*F (75-50)

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

$4

F

=

If PCTN occurs when the core reflood rate is greater than 1.0 inch per second APCT4 = 0.

The total potential PCT increase for the non-burst node is then APCT5 = APCT3 + APCT4 = 72 + 0 = 72*F Margin to the 2200*F limit is APCT6 = 2200*F - PCTN = 241*F The Fq reduction required to maintain this 2200*F clad temperature limit is (from NS-TMA-2174)

y O

.01aFq QN = (APCT AF 5 - APCT ) (10*F APCT 6

AFQN =-0 u n ess than zero.

= 0.

The peaking factor reduction required to maintain the 2200*F clad temperature limit is therefore the greater of AFqB and AF QN' AFQPENALTY "

B.

The effect on LOCA analysis results af 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 i

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 accept-able for this interim period to help offset penalties resulting from application of the NRC fuel rod models. 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 opera-tion for this interim period is determined as the appropriate AFq credit identified in Section (B) above, minus the AFQPENALTY calculated in Section (A) above (but not greater than zero).

ig ADJUSTMENT = S.28 - S.4V2

=0 CJP/4mg9A27

.

ML19309H465

Text

o

=

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