ML20148B025
| ML20148B025 | |
| 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 |
| Shared Package | |
| ML19341D844 | List: |
| References | |
| TAC-48853, TAC-48938, NUDOCS 8001160539 | |
| Download: ML20148B025 (7) | |
Text
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- L Wisconsin Electnc po.vta coursur 231 W. MICH10 AN. P.O. BOX 2046. MILW AUKEE, WI 53201 January 9, 1980 Mr. H'arold R. Der ton, Director Office of Nuclear Reactor Regulation
'U.
S. NUCLEAR REGULATORY COMMISSION Washington, D.
C.
20555 l
Dear Mr. Denton:
DOCKET NOS. 50-266 AND 50-301 FUEL CLADDING RUPTURE, STRAIN, AND FLOW f
BLOCKAGE MODELS FOR ECCS ANALYSES POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 i
Your letters dated November 9 and November 27, 1979, requested a confirmation of the representations.made on our behalf j
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 i
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 'f concern are, resolved.
o Wisconsin Electric Power Company has received from Westinghouse the results of technical evaluations of the impact.
of draft report NUREG-0630 cladding models on the most recent large-break ECCS analyses for Point Beach 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 i
ECCS analyses and the results are shown in the attachment to this letter..
.E8'1_4GT3hd.
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l-Mr. Harold R.
D$nton Jcnuary 9, 1980 1
In the' November 2, 1979 Westinghouse letter (MS-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 de'pendence factored into it.
This lack of heat-up rate dependence in-the small-break analyses of Point Beach is not a safety concern for t.he following reasons:
1.
.The "October, 1975" model contains analytical model improvements which have always resulted i
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 Epeaking factor (F ).
These interim results are extremely conservative 0
fo~r 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 hea't-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. Danton
-34 January 9, 1980 t
j Final resolution of this issue will be achieved when the differences between the fuel rod model's are resolved by Westinghouse and members of your staff.
~
Very truly yours, o
7
)A -
Executivo Vice President Sol Burstein Attachment Blind"copies to Messrs.
C.
S. McNeer R. H. Gorske/A. W.
Finke C. W. Fay D. K. Porter V G. A. Reed Gerald Charnoff 4
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pues.
I.
Evaluation of'the Potential Imoact of Usina Draft flVREG-0630 Fuel Rod
~*
Models in the Point Beach fluclear Plant (PB'tP) Loss of Coolant Accident (LOCA) Analyses A.
Previous Point Beach fluclear Plant ECCS Analyses Results The evaluation is performed on the two most limiting LOCA analyses for PBNP which are identified below:
Assumotions 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 MWt Heat Flux Hot Channel Peaking Factor (F )
2.32 2.32
~
g
. Steam Generator Tube Plugging Eighteen (18) Percent (Uniform)
Initial Reactor Coolant Pressure (psia) 2000 2280 Calculated Results l
Hot Rod Maximum Te@erature for the 1932 1929 Burst Region of the Clad (PCT )( F)
B 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 Tegerature (ft.)n)('F) 7.5 7.5 Clad Strain at the End of Blowdown
- 1. 3.
1.5 at this Elevation (%)
Maximum Clad Stra'in at this Elevation 4.9 4.7 Core Reflood Rate at the Time of Maximum
< 1.0
< 1.0 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 ave. age rod burst was predicted to occur)
B.
Evaluation of the Maximum Potential ~ Impact on the Burst Node Peak Clad Temperature for PBr4P The maximum potential impact on the peak clad temperature of tht: hot rod burst node is evaluated in. terms of a core peaking factor (F penalty required to maintain the peak temerature below 2140*F (o)
PBilP has an interim penalty of 60F on the PCT limit pending final resolution of the upper plenum injection issue).
The method of evaluation is fully explained in Westinghouse letter NS-TMA-2174 dated December 7,1979. This method reduces the Fo to maintain the PCT below the.PBNP limit of 2140*F using the following bases from the letter:
+0.01 aFg~on g 'ner'ic sensitivity studies);
150 F APCTg 1.
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2.
Use of the NRC Burst Model could require a maximum Fq reduction of 0.015;
' 3.
Use of the NRC Strain Model could require a maximum Fg reduction ~ of 0.03.
The calculation for the two Point Beach analyses is performed as follows:
APCT1 = the maximum PC.T penalty on the hot rod t!urst node
= maximum total Fo reductions converted to PCT penalty
= (0.015 +0.03)(150*F APCTB/.016F )
Q
= 675'F APCT2 = the hot rod burst node PCT margin to the PBNP limit of 2140*F
= 2140*F - PCTB
= 2140*F - 1929*F (Unit 2)
= 211*F (Unit 2) or 208*F (Unit 1)
B reduction required to maintain the PCT of the burst AF
=F 0
kodebelow2140'F
= (APCT1 - APCT
= (675*F - 211*2)((.01 AFg/150*F APCT )
F).01/150 F) (Unit 2)B
~
=.04 (Unit 2 or Unit 1)
Therefore, the maximum potential i@act of using 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 o.f 2140*F.
C.
Evaluation of the Maximum Potential Impact on the Non-Burst Hode Peak Clad T_emperature for PBitP 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 first step evaluates the i@act on the PCT of the NRC clad burst and strain models. on the pellet-clad gap conductance prior to burst, i.ower calculated strain with the use of the NRC 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 PBNP:
= (Maximum strain - blowdown strain) 20'F aPCT
.01 astrain (Unit' 2)
= (.047
.015)(20'F/.01)(Unit 1)
= $4*F (Unit 2) or 72*F,
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g...
/
The second step evaluates the impact of the'llRC burst and fuel assembly flow blockage curves on the calculated PCT. Sin.ce the maximum flow blockage indicated by the llRC curve is 75 perceni., the potential PCT increase is calculated by increasing the currently calculated flow blockage to 75. percent.
" PCT sensitivity formula based on generic sensitivity studies, which
- explained in Westinghouse letter tis-Tl4A-2174 dated December 7,19u, 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 blo'ckage) 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 PBriP.
total impact on PCT of both steps aPCT S = APCT3 + APCT4
=
= 64*F + 121*F (Unit 2)
= 185'F (Unit 2) or 193*F (Unit 1)
The core peaking factor (F ) reduction required to maintain the PCT O
less than the PBliP limit of 2140*F is calculated using another formula from letter tis-TMA-2174 as shown below:
AFh3=F reduction required to maintain the hot rod non-burst clad kemperature less than 2140 F
= (PCTrp+ APCTb-2140'F)
[.01AFn
)
\\ 10"F aPCT /
. 01 ')
(Unit 2).
= (2053*F + 185*F - 2140*F) (TFF)
(
=.10 (Unit 2) or.115 (Unit 1)
II.
The Minimum Potential Impact on LOCA Analyses Results of Using imoroved Anaivtical Models The effect on LOCA analyses results of using improved analytical and rmdeling techniques in the SATAfl blowdown computer code has been analyzed. The results were submitted to the NRC staff, for review.
4 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 (F ) of +0.12 for two loop Westinghouse plants such q
as PBitP.
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III.
Reauired Adjustment in lleat Flux flot Channel Peaking Factor (Fn)
~
The hot channel factor adjustment required to meet the PCT limit of 2140*F for PBilP is the allowable credit from Section II minus the raxic.um penalty from Sections I.B (the burst node) or I.C (the non-burst node):
q enalty =.12 - Maximum (.04 or.115), but not greater than zero aF p
" 9.
Therefore, no adjustment in Fg is required for either unit at PBNP.
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