ML19211D135

Responds to NRC 791109 & 27 Ltrs Requesting Confirmation of Representations Re Fuel Cladding & Fuel Assembly Blockage Models Used in ECCS Analyses.Final Resolution Will Be Achieved When Fuel Rod Models Are Resolved

ML19211D135
Person / Time
Site:	Point Beach
Issue date:	01/09/1980
From:	Burstein S WISCONSIN ELECTRIC POWER CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
TAC-48853, TAC-48938, NUDOCS 8001160539
Download: ML19211D135 (7)

Text

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Wisconsin Electnc eowca coursur 231 W. MICHIGAN, P.O. BOX 2046. MILWAUKEE. WI 53201 January 9, 1980 Mrs H'arold R. Denton, Director Office of Nuclear Reactor Regulation U. S. NUCLEAR REGULATORY COMMISSION Washington, D. C. 20555

Dear Mr. Denton:

DOCKET NOS. 50-266 NND 50-301 FUEL CLADDING RUPTURE, STIN, AND FLOW BLOCKAGE MODELS FOR ECCd ANALYSES POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 Your letters dated November 9 and November 27, 1979, requested a confirmation of the representations made on our behalf by Westinghouse Electric Corporation concerning the fuel cladding rupture, rupture strain, and fuel assembly flow blockage models used in our ECCS analyses, in view of the data and models presented in the draft report NUREG-0630, Cladding Swelling and Rupture Models for LOCA Analysis. These representations were made in a Westinghouse letter (NS-TMA-2147) dated November 2, 1979, and were subsequently revised by letters NS-TMA-2158 and -2163 dated November 16, 1979, and NS-TMA-2174 dated December 7, 1979.

Discussions of these representations took place on November 13, December 6, and December 20, 1979, between representatives of Westinghouse and your Staff. Methods for calculating interim penalties were agreed upon by your Staff in those discussions.

In addition, interim benefits to the analyses results which could be taken into account for recently submitted improvements to the Westinghouse large-break evaluation model were also agreed upon by your Staff. The evaluation of these ECCS analytical model considerations provided in the attachment demonstrate that plant operation may continue until differences between the fuel rod, models of concern are resolved.

Wisconsin Electric Power Company has received from Westinghouse the results of technical evaluations of the impact of draf t report NUREG-0630 cladding models on the most recent large-break ECCS analyses for Point Eeach Nuclear Plant. These analyses assume eighteen percent steam generator tube plugging and reactor coolant system operation at both 2000 and 2280 psia and the results are applicable to the current operating modes of Point Beach Nuclear Plant Units 1 and 2, respectively. This evaluation conservatively applied the penalties and benefits to the existing ECCS analyses and the results are shown in the attachment to /hCf37 this letter. 3

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Mr. Harold R. Denton January 9, 1980 In the November 2, 1979 Westinghouse letter (NS-TMA-2147) to you, it was stated that heat-up rate dependence was already factored into small-break LOCA analyses. The small-break LOCA analyses for Point Beach Nuclear Plant were performed using the

" August, 1974" Westinghouse small-break evaluation model, which does not employ heat-up rate dependent fuel rod burst curves.

The " October, 1975" model is the model which has heat-up rate dependence factored into it. This lack of heat-up rate dependence in the small-break analyses of Point Beach is not a safety concern for the following reasons:

1. The " October, 1975" model contains analytical model improvements which have always resulted in a reduction of the calculated peak clad temperature (PCT) in other Westinghouse plants over that calculated by the " August, 1974" model. This would also be the case for Point Beach.

2. The results of the Point Beach small-break analyses show that no hot rod burst occurs and that PCT is only 1367'F so that the large-break LOCA is always the limiting LOCA for ECCS evaluation.

Only the limiting large-break ECCS analyses, therefore, need to be re-evaluated, as described above.

The results of the evaluations demonstrate that both units of Point Beach continue to meet all of the ECCS acceptance criteria of 10 CFR 50.46 without any reduction in the heat flux hot channel peaking factor (F g). These interim results are extremely conservative for the following reasons:

1. The penalties assessed are maximum potential values, and the benefits allowed are minimum values.

2. A hot fuel assembly flow blockage of 75% was unrealistically assumed where 0% blockage was calculated previously for Point Beach.

(The average hot assembly rod was not calculated to burst.)

3. The Westinghouse heat-up rate dependent burst curves were used for an additional ECCS evaluation of Point Beach, and the results showed no increase in the PCT (Point Beach was Plant No. 18 in Westinghouse letter NS-TMA-2163 dated November 16, 1979).

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Mr. Harold R. Denton January 9, 1980 Final resolution of this issue will be achieved when the differences between the fuel rod models are resolved by Westinghouse and members of your staff.

Very truly yours,

^ _':d "

.-% t /) -

Exe utive Vice President Sol Burstein Attachment 175iuS9

. . ATTACHMENT I. Evaluation of the Potential Imnact of Usino Draft flUREG-0630 Fuel Rod Models in the Point Beacn iluclear Plant (PBilP Loss of Coolant Accident LOCA Analyses A. Previous Point Beach Nuclear Plant ECCS Analyses Results The evaluation is performed on the two most limiting LOCA analyses for PBNP which are identified below:

Assumptions Unit 1 Unit 2 Break Type and Location Double-Ended Cold Leg Guillotine Westinghouse ECCS Evaluation Model "Feb rua ry , 1978" Break Discharge Coefficient 0.4 0.4 Initial Core Power 102 Percent of 1518.5 telt Heat Flux Hot Channel Peaking Factor (Fq ) 2.32 2.32 Steam Generator Tube Plugging Eighteen (18) Percent (Uniform)

Initial Reactor Coolant Pressure (psia) 2000 2280 Calculated Results Hot Rod Maximum Temperature for the 1932 1929 Burst Region of the Clad (PCTB )( F)

Hot Rod Burst Elevation (ft.) 5.75 5.75 Hot Rod Maximum Temperature for 2062 2053 Non-Ruptured Region of the Clad (PCT Elevation of Maximum Temperature (ft.)N)( 7.5 F) 7.5 Clad Strain at the End of Blowdown 1.3 1.5 at this Elevation (%)

Maximum Clad Strain at this Elevation 4.9 4.7 Core Reflood Rate at the Time of Maximum < l.0 ^. . < l.0 Temperature (inches /second)

Core Reflood Heat Transfer Mode at " Steam Cooling" the Time of Maximum Temperature Hot Assembly Flow Blockage (%) 0.0 0.0 (No hot assembly average rod burst was predicted to occur)

B. Evaluation of the Maximum Potential Imcact on the Burst Node Peak Clad Temoerature for PBilP The maximum potential impact on the peak clad temperature of the hot rod burst node is evaluated in terms of a core peaking factor (F penalty required to maintain the peak temperature below PBit 2140 F (g) P has an interim penalty of 60 F on the PCT limit pending final resolution of the upper plenum injection issue). The method of evaluation is fully explained in Westingnouse letter NS-TMA-2174 dated December 7, 1979. This method reduces the Fn to maintain the PCT below the PBNP limit of 2140 F using the following bases from the letter:

1. +0.01 AFg ~ s ,150*F aPCTg (based on generic sensitivity studies);

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2. Use of the flRC Burst tiodel could require a maximum F reduction of 0.015;. 0

3. Use of the flRC Strain 14adel could require a maximum Fg reduction of 0.03.

The calculation for the two Point Beach analyses is performed as follows:

APCT1 = the maximum PCT penalty on the hot rod burst node

= maximum total Fo reductions converted to PCT penalty

= (0.015 +0.03)(150*F APCTB /.01aFQ )

= 675*F APCT2 = the hot rod burst node PCT margia to the PBilP limit of 2140 F

= 2140*F - PCTB

= 2140*F - 1929*F (Unit 2)

= 211*F (Unit 2) or 208 F (Unit 1)

B AF =F reduction required to maintain the PCT of the burst 0 kodebelow2140*F

= (aPCT 1 - APCT2 ) (.01 AFg/150 F APCT

= (675 F - 211*F)(.01/150 F) (Unit 2)B)

= .04 (Unit 2 or Unit 1)

Therefore, the maximum potential impact of usino the flRC fuel rod models for .the hot rod burst node PCT is to require a core peaking factor reduction of .04 to maintain the PCT below the PBNP limit of 2140 F.

C. Evaluation of the f4aximum Potential Imoact on the flon-Burst Node Peak Clad Temperature for PBi4P The maximum potential impact on the peak clad temperature of the hot rod non-burst node, which is located above the burst node and occurs during the reflood phase of the LOCA, is evaluated in two steps. The f1rst step evaluates the impact on the PCT of the llRC clad burst and strain models on the pellet-clad gap conductance prior to burst. Lower calculated strain with the use of the flRC models could result in increased gap conductance and higher clad temperatures. Since the maximum strain calculated with the use of the NRC models is identical to the original strain calculated during the blowdown phase of the accident, the maximum potential impact is evaluated by using the difference between the maximum and the blowdown strains. This evaluation assumes a 20 F increase in PCT per percent decrease in strain at the location of the PCT, based on several generic sensitivity studies.

The calculation is shown below for PBilP:

APCT3 = the maximum PCT penalty on the hot rod, non-burst node prior to rod burst

= (flaximum strain - blowdown strain) 20 F APCT

.01 astrain

= (.047 .015)(20 F/.01) (Unit 2)

= 44 F (Unit 2) or 72 F (Unit 1) 17:(  :

The second step evaluates the impact of the flRC burst and fuel assembly flow blockage curves onthe calculated PCT. Since the maximum flow blockage indicated by the flRC curve is 75 percent, the potential PCT increase is calculated by increasing the currently calculated flow blockage to 75 percent. A PCT sensitivity formula based on generic sensitivity studies, which was explained in Westinghouse letter tis-TMA-2174 dated December 7,1979, is used for the PBriP calculation as shown below:

APCT4 = the maximum PCT penalty on the hot rod, non-burst node following rod burst

= 1.25 F APCT (50% - Percent current blockage) 1% ablockage

+ 2.36*F APCT (75% - 50%)

1% ablockage

= 1.25*F (50% - 0%) + 2.36 F (75% - 50%) (Unit 2)

= 121 F (Unit 2 or Unit 1) flote: If core reflood rate is greater than 1.0 inches /second, then APCT4 = 0. This is not applicable to PBt1P.

APCT total inpact on PCT of both steps S == APCT3 + APCT4

= 64*F + 121*F (Unit 2)

= 185*F (Unit 2) or 193*F (Unit 1)

The core peaking factor (FO ) reduction required to maintain the PCT less than the PBi1P limit of 2140 F is calculated using another formula from letter i4S-TMA-2174 as shown below:

AF[B=F reduction required to maintain the hot rod non-burst clad kemperature less than 2140 F

= (PCT;r + aPCTS - 2140 F) [.01AFo 3 (10"F APCT /

= (2053 F + 185*F - 2140 F) [.01 ) (Unit 2)

\ 10"F) .

= .10 (Unit 2) or .115 (Unit 1)

II. The Minimum Potential Imoact on LOCA Analyses Results of Using Improved Analytical Mocels The effect on LOCA analyses results of using improved analytical and modeling techniques in the SATAtl blowdown computer code has been analyzed. The results were submitted to the flRC staff, for review.

An initial review of those results by the staff has allowed the establishment of a credit to offset the penalties for the interim period. This credit is an increase in the allowable heat flux hot channel factor (Fg) of +0.12 for two loop Westinghouse plants such as PBriP.

\159\D2 III. Reouired Adjustment in Heat Flux flot Channel Peakina Factor (Fo)

The hot channel factor adjustment required to meet the PCT limit of 2140*F for PBitP is the allowable credit from Section II minus the maximum penalty from Sections I.B (the burst node) or I.C (the non-burst node):

AFg penalty = .12 - Maximum (.04 or .115), but not greater than zero "E

Therefore, no adjustment in Fg is required for either unit at PBf1P.

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