ML20140A775
| ML20140A775 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 12/31/1985 |
| From: | WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20140A716 | List: |
| References | |
| NUDOCS 8601230369 | |
| Download: ML20140A775 (11) | |
Text
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EXHIBIT C PRAIRIE ISLAND NUCLEAR GENERATING PLANT License Amendment Request Dated January'13, 1986 DEMONSTRATION OF THE CONFORMANCE OF PRAIRIE ISLAND UNITS WITH EXXON FUEL ASSEMBLIES TO APPENDIX K AND 10CFR50.46 FOR SMALL BREAK LOCAs Westinghouse Electric Corporation Nuclear Technology Division Nuclear Safety Department Safeguards Engineering and Development December 1985 8601230369 860113 '
PDR ADOCK 05000292 P
PDR C-1
1 I.
Introduction This document reports the results of an analysis that was performed to demonstrate that Prairie Island, Units I and II with Exxon fuel i
casamblies, meet the requirements of Appendix K and 10CFR50.46 for I
cua11 break LOCA.
The analysis incorporates anticipated plant hardware nodifications, i.e.,
the new upper reactor internals package and the fuel assembly thimble plug removal, as well as increased levels of Fq to 2.5 and 10% steam generator tube plugging.
II. Method of Analysis The analysis was performed for Exxon fuel assemblies in the Prairie Island Units and was an extension of an analysis for small breaks with Westinghouse fuel in the same Units.
Exxon fuel parameters which effect hydraulic performance are similar to those of Westinghouse fuel, with only 3.4% greater core pressure loss and 6% less core flow area.
Given the significant similarity in hydraulic characteristics and the limited impact of core hydraulics during the quasi-steady state small break I4CA transient, the Westinghouse fuel NOTRUMP calculations cdequately defined the hydraulic transient for the Exxon fuel.
- Thus, the fuel rod response calculation using the Exxon fuel parameters in the LOCTA code was performed with the NOTRUMP inputs from the Westinghouse fuel analysis.
The initial analysis for Westinghouse fuel used the W NOTRUMP and LOCTA computer codes for a spectrum of small break sizes.
These codes are incorporated in the approved Westinghouse ECCS Small Break Evaluation Model developed to determine the RCS response to design basis small break LOCAs and to address the NRC concerns exprePt4* in NUREG-0611,
" Generic Evaluation of Feedwater Transients and St-k.
Break' Loss-of-j Coolant Accidents in Westinghouse-Designed Operati).; Plants."
Analysis
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of the Westinghouse fuel showed that the 4 inch break sizs is limiting for the Prairie Island Units, as the 3 inch and 6 inch Ursaks showed no core uncovery.
Conseguently, the Exxon fuel analysis was performed for only the 4 inch break size.
The NOTRUMP computer code is a one-dimensional general network code consisting of a number of advanced features, including the calculation of thermal non-equilibrium in all fluid volumes, flow regime-dependent drift flux multiple-stacked fluid nodes, and regime-dependent heat transfer correlations.
NOTRUMP includes the representation of the reactor core as heated control volumes with an associated bubble rise model to per=it a transient mixture height calculation.
The multinode capability of the program enables an explicit and detailed spatial representation of various system components.
In particular, it enables a proper calculation of the behavior of the loop seal during a loss-of-coolant transient.
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G-2 a
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1 Cladding thermal analyses are performed with the I4CTA-IV (Reference 3) code which uses the RCS pressure, fuel rod power history, steam flow i
past the uncovered part of the core, and mixture height history from the NOTRUMP hydraulic calculations, as input.
In this evaluation, l
Exxon fuel rod parameters were input to the LOCTA code.. The fuel design parameters were prepared by the Westinghouse Nuclear Fuel i
Division using NRC approved methodology and fuel performance models, j
modified to describe measured Exxon fuel operating performance data.
The similiarity of Exxon and Westinghouse fuel as-built and irradiated i
mechanical properties supports the validity of the model development.
l The LOCA fuel design parameters calculated by Westinghouse were then compared to with LOCA fuel parameters used by Exxon in the previous cycle LOCA evaluation.
This comparison showed good agreement on the fuel temperatures and stored energy and somewhat lower fuel rod internal pressures as a function of fuel rod power.
To assure that the new LOCA fuel performance parameters conservatively bound the values used by Exxon in the prior cycle analysis, the values calculated with the Westinghouse models were adjusted upward to match the prior cycle i
limiting values.
The fuel parameters, calculated with the modified models, were used as input to the LOCTA calculations.
Table 1 lists important input parameters and initial conditions used in the NOTRUMP analysis. The core power decay and axial power distribution are shown in Figures 1 and 2.
For these analyses, the SI delivery considers pump injection flow which is depicted in Figure 3 as a function of RCS pressure.
Minimum safeguards Emergency Core Cooling j
System capability and operability has, also, been assumed in this j
analysis.
]
The hydraulic analyses are performed with the NOTRUMP code using 102 percent of the licensed NSSS core power.
The core thermal transient analyses are performed with the LOCTA-IV code using 102 percent of i
licensed NSSS core power.
III.
Results and Conclusions j
The 4-inch break shows two brief periods of core uncovery, Figure 5, prior to accumulator injection.
During the second period of uncovery a C
i PCT of 978 F occurs, Figure 6.
This value is well below all Acceptance i
Criteria limits of 10CFR50.46 and is non-limiting in comparison to large break analysis results.
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REFERENCES i
1.
- Lee, H.,
Rupprecht, S.
D.,
Tauche, W.
D.,
- Schwarz, W.
R.,
" Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," WCAP-10054-P-A, August 1985.
I 2.
Meyer, P.
E., "NOTRUMP, A Nodal Transient Small Break and General Network Code," WCAP-10079-P-A, August 1985.
1 3.
Bordelen, F. M., et al., "IACTA-IV Program:
Loss-of-Coolant Transient Analysis," WCAP-8301 (Proprietary), and WCAP-8305 (Non-Proprietary), June 1974.
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TABLE 1 INPUT PARAMETERS USED IN TME SMALL BREAK ANALYSES Parameter Small Break Peak Linear Power (kw/ft) 15.03 (includes 102% factor)
Total Peaking Factor, FQ 2.50 Power Shape See Figure 2 Fuel Assembly Array 14x14 OFA Nominal Cold Leg Accumulator 1266 Water Volume (ft3/ accumulator)
Nominal Cold Lag Accumulator 2000 Tank Volume (ft3/ accumulator)
Minimum Cold Lag Accumulator 715 Gas Pressure (psia)
Pumped Safety Injection Flow See Figure 3 i
Steam Generator Initial Pressure (psia) 733.0 Steam Generator Tube Plugging Level %
10 Fuel Assembly Thimble Plugs Removed Reactor Upper Internals Package New Design l
l l
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C-$
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19632-3 100_
~
7 5
2 TOTAL RESIDUAL HEAT (WITH 47; S M COWN)
Y 10- I _-
E 7
- P q 5 t
^
<g2 e
h !O-2,_
l
- 7 O i
5 2
''8I
' I I ' ' ' '!
' ' ' ' l l ' I!
' 8
'I
l ic-3 IC-3 2 5
i00 2 5
10' 2
5 102 2 5
103 2 5
10' TIME AFTER Sr4JTDOWN (SEOONOS)
Figure 1. Core Power After Reactor Trip G-6
i 15632 2 16 14 12 k 10 k5 C
6 c.
8e 6
9 4
2 -
l I
I O
O 2
s 12 CORE HEIGHT (FT Figure 2. Small Sreak Power Oistribution Assumed for Loca Analysis G-7
o 15532-4 2400 2200 2000 1800
.to 1600 W
C.
~
l400 w
s 1200 E
c.
i000 2w
&W 800 u) 600 400 200 O
O 100 200 300 400 FLOW (LS/SEC)
Figure 3. Safety injection Flowrate Versus Pressure i
l G-8
e e
15532 7 L
2500 2000 R
X 1500 N
aC c toco -
OW 500 3
I I
I I
i 0
C 200 400 600 000 oco
]
TIME (SECONDS) i Figure 4. 4. Inch Cold Leg Break RCS Pressure Versus Time G-9
e e
15532s 50 40
.u
$ 30
- .c u
m
_ _ _ _ _ _ _ _ _ _7_0_2 C_F O_C_R_E_ _ _ _ _ _ _ _ _.
L 20 s
,.=
R wu IO I
I i
i t
i i
i I
0 0
200 400 600 000 1000 TIME (E,LONOS)
Figure 5. 4. Inch Cold Leg Break Core Mixture Level Versus Time G-10
e
't 15532-9A 3000 C.
~
2500 S 2000 W3
<c: isOO e
g 000 5
1 W
L 500 g
f f
f f
O O
100 200 300 400 500 TIME (SECorCS)
Figure 6. 4 inch Cold Leg Break Clad Average Temperature, Hot Rod, Versus Time G-11