ML20140A747
| ML20140A747 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 01/31/1986 |
| From: | WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20140A716 | List: |
| References | |
| NUDOCS 8601230361 | |
| Download: ML20140A747 (25) | |
Text
d EXHIBIT E PRAIRIE ISLAND NUCLEAR GENERATING PLANT License Amendment Request Dated January 13, 1986 DEMONSTRATION OF THE CONFORMANCE OF PRAIRIE ISLAND UNITS TO APPENDIX K AND 10CFR50.46 FOR LARGE BREAK LOCAs r.
Westinghouse Electric Corporation Nuclear Technology Division Nuclear Safety Department Safeguards Engineering and Development January 1986 8601230361 860113 i
PDR ADOCK O y2 -
E-1 P
i
I.
Introduction This document reports the results of an analysis that was performed to demonstrate that Prairie Island, Units I and II, meet the requirements of Appendix K and 10CFR50.46 for Large Break Loss-of-Coolant-Accidents (LOCA).
The analysis incorporates anticipated plant hardware modifications, i.e.,
the new upper reactor internals package and the i
thimble plug removal.
II. Method of Analysis The analysis was performed using the Westinghouse 1981 Evaluation Model (Raference 6) for a spectrum of break coefficients.
The Westinghouse 1981 ECCS Large Break Evaluation Model was developed to determine the RCS response to design basis large break lOCAs (see References 6-13,15,16).
Tho hydraulic analyses and core thermal transient analyses were performed with the 1981 Evaluation Model code using 102 percent of licensed NSSS core power.
The 1981 Evaluation Model is comprised of the SATAN-VI, WREFLOOD,' COCO, and LOCTA IV computer codes (References 2,3,4 and 1, respectively, see also Reference 8).
The SATAN-VI code was used to ganerate the blowdown portion of the transient, the WREFLOOD CODE was ussd to generate the refill /reflood system hydraulics, and the COCO code was used to evaluate the containment response.
Cladding thermal analyses ware performed with the LOCTA-IV code which uses the RCS pressure, fuel rod power history, steam flow past the uncovered part of the core, and mixture height history from the SATAN-VI and WREFLOOD codes as input.
The fuel parameters used as input for the LOCA analysis were generated using the Revised Thermal PAD Model.
Due to the use of the Revised Thermal PAD Model, Westinghouse has evaluated the effect of burnup on psak cladding temperatures (PCT) predicted for the Loss-of-Coolant Accident through the maximum burnup level of cycle 11 using the currently cpproved LOCA models (1981 EM) as required by Reference 18.
At a burnup of 22,000 MWD /MTU (maximum burnup for either Unit during cycle 11), the 0
burnup evaluation predicted a PCT of 1934 F compared with a PCT of 0
2098 F for the Beginning-of-life (maximum densification) case, damonstrating that the time of maximum densification remains limiting in terms of peak clad temperature.
Table 1 shows the time sequence of events for the Large Break LOCA trcnsients.
Table 2 provides a brief summary of the important results of the LOCA analyses for each case.
Figures 1 and 2 show important core characteristics during the blowdown phase of the transient (Core Pressure and Core Flow versus Time, respectively).
Figures 3 and 4 indicate the flow of ECCS water into the RCS (Accumulator Flow and Pumped ECCS Flow versus Time, respectively).
The flooding rate during the reflood portion of the transient are given in Figure 5.
Clad Average Terparatures as a function of time indicating peak clad temperatures are given in Figure 6.
The Safety Injection (SI) system was assumed to be delivering to the RCS 22 seconds after the generation of a safety injection signal.
The 22-cecond delay includes time required for diesel startup and loading of tha safety injection pumps onto the emergency buses.
Minimum safeguards Emargency Core Cooling System capability and operability ham, also, been accumed in this analysis.
E-2
Three break size coefficients were evaluated;CD = 0.4, CD = 0.6, and CD = 0.8.
These transients were considered to be terminated when the hot rod clad average temperature " turned around" (i.e. - hot rod clad cvorage temperature began to decline) indicating that the peak clad tcmperature had been reached.
III.
Results and Conclusions of the three break sizes evaluated, the Co = 0.4 break proved to be the 0
limiting (highest PCT) case with a peak clad temperature of 2098 F, 0
0 J
compared with PCTs of 2000 F and 1999 F for the CD = 0.6 and CD = 0.8 cases, respectively.
Current NRC restrictions require that a i
pannity be assessed and imposed for insufficient modeling of upper plenum ingcction(References 14,16,20,21).
This penalty was assessed to be 0
78 F for the limiting case.
In addition, a penalty of 10 F was imposed to account for hydaulic mismatch (crossflow) in the transition l
coro.
Imposition of these penalties results in a final peak clad 0
0 temperature of 2186 F which is below the 2200 F Acceptance Critaria limit established by Appendix K of 10CFR50.46 (Reference 2).
l O
E-3
REFERENCES i
1.
Bordelon, F.
M.,
et al., LOCTA-IV Procram: Loss-of-Coolant i
Transient Analysis, WCAP 8301 (Proprietary Version), WCAP 8305 (Non-Proprietary Version), June 1974.
2.
" Acceptance criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors: 10CFR 50.46 and Appendix K of 10CFR 50.46," Federal Reaister, Vol. 39,No.
3, January 4, 1974.
3.
Bordelon, F.
M.,
et al., SATAN-VI Procram: Comorehensive Snace-Time Denendent Analysis of Loss-of-Coolant, WCAP 8302 (Proprietary Version), WCAP 8306 (Non-Proprietary Version), June i
1974.
4.
Kelly, R.
D.,
et al., Calculational 1[odel for Core Refloodina after a Loss-of-Coolant Accident (WREFLOOD Codel, WCAP 8170 (Proprietary Version), WCAP 8171 (Non-Proprietary Version), June 1974.
5.
Bordelon, F.
M.,
and E. T. Murphy, Containment Pressure Analysis Code (COCO), WCAP 8327 (Proprietary Version), WCAP 8326 (Non-Proprietary Version), June 1974.
6.
Eicheldinger, C., Westinchouse ECCS Evaluation Model. 1981 l
Version, WCAP 9220-P-A (Proprietary Version), WCAP 9221-A (Non-Proprietay Version), Rev.
1, 1981.
7.
Bordelen, F.
M.,
H. W. Massie, and T. A. Zordan, Westinahouse ECCS Evaluation Model-Summarv, WCAP 8339, July 1974.
8.
Bordelon, F.
M.,
et al., The Westinchouse ECCS Evaluation Model:
Sucolementary Informat[2D, WCAP 8471 (Proprietary Version, WCAP 8472 (Non-Proprietary Version), January 1975.
9.
Salvatori, R., Westinnhouse ECCS - Plant Sansitivity Studies, WCAP 8340 (Proprietary Version), WCAP 8356 (Non-Proprietary Version), July 1974.
10.
Delsignore, T.,
et al., Wagtinchouse ECCS Two-Loco Sensitivity Studies (14 x 14), WCAP 8854 (Non-Proprietary Version), September 1976.
i 1
i E-4
~
11.
Westinchouse ECCS Evaluation Model Sensitivity Studies, WCAP 8341 (Proprietary Version), WCAP 8342 (Non-Proprietary Version), 1974.
12.
Kelly, R.
D.,
C. M. Thompson, et al., Westinchouse Emercancy Core Coolina System Evaluation Model for Analyzina Laraa LOCAs Durina Ooeration With One Loon Out of-Service for Plants Without LooD Isolation Valves, WCAP 9166, February 1978.
13.
Safety Evaluation Recort on ECCS Evaluation Model for Westinnhouse Two-Loon Plants, November 1977.
14.
"NRC Questions Regarding the January 16, 1978 submittal by Westinghouse Designed Two-Loop Plant Operators," February 1, 1978.
15.
Letter from T. M. Anderson, Westinghouse Electric Corporation, to J. Stolz, NRC, (NS-TMA-1830), dated June 1978.
16.
Letter from T. M. Anderson, Westinghouse Elastric Corporation, to J. Stolz, NRC, (NS-TMA-1834), dated June 20, 1978.
17.
" Safety Evaluation Report on Interim ECCS Evaluation Model for Westinghouse Two-Loop Plants," March 1978.
18.
Letter from Cecil O. Thomas (NRC) to E. P. Rahe, Jr.
(Westinghouse), " Acceptance for Referencing of Licensing of Topical Report WCAP-8720, Addendum 2,
' Revised PAD Code Thermal Safety Model'", Dated December 9, 1983.
19s
" Westinghouse Revised PAD Code Thermal Safety Model," WCAP-8720, Addendum 2 (Proprietary), and WCAP-8785 (Non-Proprietary).
i 1
E-5
TABLE 1 i
LARGE BREAK TIME SEQUENCE OF EVENTS 4
DECLG (CD = 0.8)
DECLG (CD = 0.6)
DECLG (CD = 0.4)
(Sec)
(Sec)
(Sec)
Start 0.0 0.0 0.0 R0 actor Trip
.568
.576
.590 Signal S.I. Signal
.47
.53
.64 q
Acc. Injection 5.12 6.58 9.41 End of Blowdown 16.57 18.43 22.56 Pump Injection 22.47 22.53 22.64 Bottom of Core 29.11 31.14 35.39 Recovery Acc. Empty 38.8 40.7 44.3 O
E-6 i
TABLE 2 LARGE BREAK RESULTS DECLG (CD = 0.8)
DECLG (CD = 0.6)
DECLG (CD = 0.4)
Paak Clad Temp., oF
- w/ penalties 2087.
2088.
2186.
Paak Clad Temp., CF 1999.
2000.
2098.
Paak Clad Temp.
7.5 7.5 7.5 j
Location, ft Local Zr/
3.840 3.827 6.284 H O Rxn (max), %
2 Local Zr/
7.5 7.5 7.5 H O Location, ft 2
Total Zr/H O Rxn, %
<0.3
<0.3
<0.3 2
Hot Rod Burst Time, 70.6 72.0 70.8 cec Hot Rod Burst 7.0 7.25 6.5 Location, ft Cciculation:
NSSS Power, MWt, 102% of 1650 P32k Linear Power, kw/ft, 102% of 14.25 Paaking Factor (At Design Rating) 2.30 Accumulator Water Volume (Cubic l
Feet per Tank-Nominal) 1266.5 Accumulator Pressure, psia 715 Number of Safety Injection Pumps Operating 3
Stcam Ga.v rator Tubes Plugged 5%
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FIGURE 6b PEAR CLAD TEMPERATURE - DECIA (CD=0.6) 4 4
g Nort: Asterisks (*) do D91 Y*Present a separate curve, but provide a tracer to identify the curve associated with the peak (highest PCT) node.
Where peak and burst nodes coincide, only one curve (with asterisk tracer) will be seen.
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8 NSP NEW UI 5% TUBE PLUGGING CD=0.4 DECLG FQ=2.30 CLAD AVG. TEMP. HOT ROD
- BURST, 6.50 FT( )
PEAK, 7.50 FT(*)
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a se is see its ese its aos ars no ars see sn 12/03/85 FIGURE 6c PEAR CIAD TEMPERATURE - DECIA (CD=0.4)
NOTE: Asterleks (*) do not represent a esperate curve, but provide a tracer to identify the curve associated with the peak (highest PCT) node.
Where peak and burst trwlas coincide, only one curve (with asterlek tracer) will be seen.