ML20024A901
| ML20024A901 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 05/27/1983 |
| From: | Chandler J, Kayser W, Tahvili T SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML20024A891 | List: |
| References | |
| XN-NF-83-38, NUDOCS 8307010157 | |
| Download: ML20024A901 (56) | |
Text
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XN-NF-83-38 Issue Date: 5/27/83 PRAIRIE ISLAND UNITS 1 AND 2 LIMITING BREAK LOCA-ECCS ANALYSIS USING EXEM/PWR 1
l I
i Prepared by:
7 7M' S/ 5/' 83 l
T. Tahvili, Project Manager PWR Safety Analysis Concur:
Mk MS 3
l W. V. Ka'yser, Manager PWR Safety Analysis f!f
.2 Concur:
J. C. Charidler Reload Fuel Licensing Approve:
4'MA4f3 R. B. Stout, Manager Licensing & Safety Engineering Approve:
~
h 9M A 783 G. A. Sofer, Manager Fuel Engineering & Rchnical Services t
I gf ERON NUCLEAR COMPANY,Inc.
8307010157 830624 PDR ADOCK 05000282 P
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NUCLEAR REGULATORY COMMISSICN DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY 1
1 This technical report was eterived through research and development programs sponsored by Exxon Nuclear Company, Inc. It is being sub-mitted by Exxon Nuclear to the USNRC as part of a technical contri-f bution to facilitate safety analyses by licensees of the USNRC which 1
utilize Exxon Nudeer fabricated reload fuel or other technical services i
provided by Exxon Nuclear for licht water power reactors and it is true and correct to the best of Exxon Nuclear's knowledge, information, and belief. The information contained herein may be used by the USNRC in its review of this report, and by licensees or applicants before the USNRC which are customers of Exxon Nuclear in their demonstration i
of compliance with the USNRC's regulations.
Without derogating from the foregoirg neither Exxon Nuclear nor any person acting on its behalf:
A. Maltas any warranty, express or implied, with respect to the accura:y, completeness, or usefulness of the infor-mation contained in this documeit, or that the use of any information, apparatus, method, or process disclosed in this document will not infringe privately owned rights; or 8.
Assumes any liabilities with respect to the use of, or Mr derrages resalting from the use of, any information, ap-paratus, met'ed, or process disclosed in this document.
XN NF-F00,766 e
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i XN-NF-83-38 TABLE OF CONTENTS Section Page
1.0 INTRODUCTION
AND
SUMMARY
1
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2.0 LIMITING BREAK LOCA ANALYSES.......................
4 j
2.1 LOCA ANALYSIS MODEL...........................
4
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2.2 -RESULTS.......................................
6
3.0 CONCLUSION
45
4.0 REFERENCES
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ii XN-NF-83-38 l
I LIST OF TABLES
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Table Page 1.1.
Prairie Island Units 1 and 2 TOPROD LOCA-ECCS Analysis Results.........................
3 2.1 Prairie Island Units 1 and 2 System Data...........
8 1
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2.2 Fue l De s i g n P a rame te r s.............................
9
-2.3 Prairie Island Units 1 and 2 TOPR00 LOCA-ECCS Analysis Resul ts, Event Times............
10 o
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iii XN-NF-83-38 LIST OF FIGURES Figure Page 2.1 RELAP4/EM Blowdown System Nodalization for Prairie Island Unit 1 and 2....................
11 2.2 Axial Peaking Factor versus Rod Length, 0.4 DECLG Break....................................
12 2.3 Downcomer Flow Rate, 0.4 DECLG Break...............
13 2.4 Upper Plenum Pressure, 0.4 DECLG Break.............
14 2.5 Average Core Inlet Flow, 0.4 DECLG Break...........
15 2.6 Average Core Outlet Flow, 0.4 DECLG Break..........
16 2.7 Total Break Flow, 0.4 DECLG Break..................
17 2.8 Flow from Intact Loop Accumulator, 0.4 DECLG Break....................................
18 2.9 Pressurizer Surge Line Flow, 0.4 DECLG Break....................................
19 2.10 Containment Back Pressure, 0.4 DECLG Break.........
20 2.11 Hot Channel Heat Transfer Coefficient, 0.4 DECLG Break, 0-15,000 MWD /MTM..................
21 2.12 Clad Surface Temperature, 0.4 DECLG Break, 0-15,000 MWD /MTM..................
22 2.13 Depth of Metal-Water Reaction, 0.4 DECLG Break, 0-15,000 MWD /MTM..................
23 2.14 Hot Channel Average Fuel Temperature, 0.4 DECLG Break, 0-15,000 MWD /MTM..................
24 l
2.15 Hot Assembly Inlet Flow, l
0.4 DECLG Break, 0-15,000 MWD /MTM..................
25 2.16 Hot Assembly Outlet Flow, 0.4 DECLG Break, 0-15,000 MWD /MTM..................
26 I
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iv XN-NF-83-38 LIST OF FIGURES (Cont.)
Figure Page 2.17 Hot Channel Heat Transfer Coefficient, 0.4 DECLG Break, E0L...............................
27
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2.18 Clad Surface Temperature, 0.4 DECLG Break, E0L...............................
28 1
2.19 Depth of Metal-Water Reaction, 0.4 DECLG Break, E0L...............................
29
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2.20 Hot Channel Average Fuel Temperature, 0.4 DECLG Break, EOL...............................
30 2.21 Hot Assembly Inlet Flow, 0.4 DECLG Break, E0L...............................
31 2.22 Hot Assembly Outlet Flow, 0.4 DECLG Break, EOL...............................
32 2.23 Normalized Power, 0.4 DECLG Break, 0-15,000 MWD /MTM...................................
33 2.24 Normalized Power, 0.4 DECLG Break, E0L.............
34 2.25 Reflood Core Mixture Level, 0.4 DECLG Break, 0-15,000 MWD /MTM..................
35 2.26 Reflood Downcomer Mixture Level, 0.4 DECLG Break, 0-15,000 MWD /MTM..................
36 2.27 Reflood Upper Plenum Pressure, 0.4 DECLG Break, 0-15,000 MWD /MTM..................
37 2.28 Core Flooding Rate, 0.4 DECLG Break, 0-15,000 MWD /MTM..................
38 2.29 Reflood Core Mixture Level, 0.4 DECLG Break, E0L...............................
39 2.30 Reflood Downcomer Mixture Level, 0.4 DECLG Break, E0L...............................
40
v XN-NF-83-38 l
l LIST OF FIGURES (Cont.)
Figure' Page 2.31 Reflood Upper Plenum Pressure, 0.4 DECLG Break, E0L..............................
41 2.32 Core Flooding Rate, 0.4 DECLG Break, E0L..........
42 2.33 T000EE2 Cladding Temperature vs Time, 0.4 DECLG Break, 0-15,000 MWD /MTM.................
43
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2.34 T00DEE2 Cladding Temperature vs Time, 0.4 DECLG Break, E0L..............................
44
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1 XN-NF-83-38
1.0 INTRODUCTION
AND
SUMMARY
This document presents analytical results for a postulated large break loss-of-coolant accident (LOCA), performed for the Prairie Island Units 1 and 2 nuclear reactors. The analyses assume a reactor operating power of 1683 MWt (includes 2% power uncertainty), and use of Exxon Nuclear Company's (ENC's)
TOPROD fuel.
The calculations were made for the double-ended cold leg l
guillotine break, with a discharge coefficient of 0.4 (0.4 DECLG) identified in the previous analyses as the most limiting break.(1,2,3)
)
The analyses were performed using the EXEM/PWR ECCS evaluation model(4),
with the R0DEX2 computer model for evaluating the rod stored energy and fission gas release (5). The EXEM/PWR ECCS evaluation model includes the NRC fuel swelling and flow blockage model, NUREG-0630.(14)
The analyses are applicabie up to a five percent (5%) steam generator (SG) tube plugging,-and maximum peak pellet exposure limit of 55,000 MWD /MTM. The allowable linear heat generation rate, including the 1.02 factor for power uncertainty, was T
15.02 kW/f t, corresponding to a total power peaking factor of 2.32 (Fg ), and nuclearenthalpyriseof1.55(F[g)fortheentireexposure.
The analyses were performed assuming an entire core with TOPROD fuel.
With respect to a LOCA, the TOPROD fuel design is more limiting than previous ENC XN-1 and XN-2 reload fuel designs in Prairie Island Units 1 and 2.
This is due to the increased core flow area which reduces core reflood rates in the LOCA analysis for TOPR00 fuel and results in higher PCTs. This analysis is l
l therefore applicable to the XN-1 and XN-2 fuel designs for peak pellet burnups less than 55,000 MWD /MTM.
i
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XN-NF-83-38 The calculational basis and results are summarized in Table 1.1. The maximum calculated peak cladding temperature (PCT) is 21420F, occurring at 196 seconds into the accident at a location 9.13 feet from the bottom of the active core, with a total metal-water reaction less than one percent. The 21420F PCT includes a 100F temperature addition due to the use of NRC interim upper plenum injection model(6) as modified by Westinghouse (7). The results of the analyses show that within the limits established, the Prairie Island Nuclear Reactors operating at the stated power level, and with steam generator
(
tube plugging up to 5%, satisfy the criteria specified by 10 CFR 50.46(8),
l
Table 1.1 Prairie Island Units 1 and 2 TOPROD LOCA-ECCS Analysis Results 0 - 15000 MWD /MTM 15000 - 55000 MWD /MTM Analysis Results Peak Pellet Exposure Peak Pellet Exposure Peak Clad Temperature (PCT), OF 2091 2142 APCT for UPI, GF
-10 10 Time of PCT, sec. 191 196 Peak Clad Temperature Location, ft.
8.88 9.13 Local Zr/H O Reaction (max.), %*
4.68 5.60 2
Local Zr/H O Location, ft. from bottom 8.63 9.13 2
Total H2 Generation,.% of total Zr Reacted
< 1.0
< 1.0 m
Hot. Rod Burst Time, sec.
30.19 30.19 Hot Rod Burst Location, ft.
6.0 6.0 Calculational Basis License Core Power, MWt 1650 l
-Power Used for Analysis, MWt**
1683 Peak Linear Power for Analysis, kW/ft**
15.02 TotalPeakingFactor,Fh2H 2.32 Enthalpy Rise, Nuclear, F 1.55 g
Steam Generator Tube Plugging (%)
5.00 k
h Computer value at 380 seconds.
M Including 1.02 factor for power uncertainties.
{
4 XN-NF-83-38 2.0 LIMITING BREAK LOCA ANALYSIS I
This report provides LOCA-ECCS analyses performed for Prairie Island Units 1 and 2 with a steam generator tube plugging up to 5%. The analytical techniques used are in compliance with Appendix K of 10 CFR 50, and are described in the ENC WREM models(9), and the Emergency Core Cooling System Evaluation Model Updates: WREM-II(17), WREM-IIA (13) and EXEM/PWR(4).
A LOCA break spectrum analysis was performed and reported in XN-NF 45(1).
The limiting LOCA break was determined to be a large double-ended guillotine break of the cold leg, with a discharge coefficient of 0.4 (0.4 DECLG). The analyses performed and reported herein for the 0.4 DECLG break consider:
(1) A revised stored energy model R0DEX2(5) in place of the previously applied GAPEX(10) model.
(2) The NRC upper plenum injection (UPI) interim model, developed by theNRCStaff(6)andmodifiedbyWestinghouse(7),
(3) Updates to the latest Prairie Island Units 1 and 2 application to reflect all model revisions and documented in XN-NF-82-20(P), Revision 1(4).
2.1 LOCA ANALYSIS MODEL The Exxon Nuclear Company EXEM/PWR ECCS evaluation model(4) was used to perform the analyses required. This model consists of the following computer codes: RODEX2(5) code for initial rod stored energy and internal fuel i
rod gas inventory; RELAP4-EM(ll) for the system blowdown and hot channel blowdown calculations; CONTEMPT-LT/22 as modified in CSB 6-1(16) for com-putation of containment backpressure; REFLEX (4.14) for computation of system reflood; and T000EE2(4,14,15) for the calculation of final fuel rod heatup.
5 XN-NF-83-38 The Prairie Island nuclear reactor is a two-loop Westinghouse pressurized water reactor with an upper plenum injection and dry containment.
The reactor coolant system is nodalized into control volumes representing reasonably homogeneous regions, interconnected by flow-paths or " junctions" as described in XN-NF-77-25(A)(16). The system nodalization is depicted in Figure 2.1.
The unbroken loops were assumed symmetrical and modeled as one intact loop with appropriately scaled input.
The pump performance charac-teristic curves are supplied by the NSSS vendor. Five percent of the steam generator tubes are assumed to be uniformly plugged. The transient behavior was determined from the governing conservation equations for mass, energy, and momentum. Energy transport, flow rates, and heat transfer are determined from appropriate correlations.
System input parameters are given in Table 2.1.
The reactor core is modeled with heat generation rates determined from reactor kinetics equations with reactivity feedback and with decay heating as required by Appendix K of 10 CFR 50. The chopped cosine axial power profile used for the analyses is shown in Figure 2.2, with a maximum axial peaking factor of 1.453, corresponding to a total peaking factor of 2.32, and Ffgof1.55. This axial power profile in configuration with the current K(Z) function developed by the NSSS vendor will not be used to define operating envelop for Fg. The analysis of the loss-of-coolant accident is performed at 102 percent of rated power. The fuel design parameters are shown in Table 2.2.
1 Two cases of LOCA-ECCS calculations were performed with input which j
bounds the fuel history up to 55,000 MWD /MTM peak pellet exposure. The most limiting fuel conditions from beginning-of-life to 15,000 MWD /MTM (first
L l
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XN-NF-83-38 i
case), and from 15,000 MWD /MTM to end-of-life (second case) were determined and used in each calculation.
Decay power, internal rod pressure and the fission gas releases were highest at E0L (second case) for the hot rod, while stored energy was calculated to be highest at lower exposure (first case).
The combination of highest stored energy, rod pressure, and decay power was used to bound the LOCA-ECCS analysis over the exposure ranges shown.
The small rod diameter for ENC TOPROD fuel, as compared to other fuel designs in the Prairie Island reactors, results in a larger core flow area.
The larger core flow area decreases the core flooding rates, which results in higher PCTs.
In addition, the 55,000 MWD /MTM exposure limit considered in this analysis encompasses the exposure limits expected for the previous ENC XN-1 and XN-2 fuel designs operating in Prairie Island units.
Therefore, the LOCA-ECCS analyses reported in this document bound the previous Prairie Island ENC fuel designs.
2.2 RESULTS Table 2.3 presents the timing and sequence of events as determined for the large guillotine break with a discharge coefficient of 0.4.
Comparison of these results with the previous LOCA-ECCS analysis for a TOPROD fuel shows very slight change in the event times.
Figures 2.3 through 2.9 present plotted results for system blowdown analysis. Unless otherwise noted on the figures, time zero corresponds to the time of break initiations.
I Figure 2.10 presents calculated containment backpressure time history.
Figures' 2.11 through 2.22 present results for the hot channel blowdown calculations.
Figures.2.23 and 2.24 show the normalized power calculation results. The reflood calculation results are shown in Figures 2.25 through 2.32.
7 XN-NF-83-38 i
The maximum peak cladding temperature (PCT) calculated for the 0.4 DECLG break at the E0L is 21420F (Figure 2.34). This value includes a 100F temperature addition associated with the use of the NRC interim upper plenum injection (UPI) model as modified by Westinghouse. The maximum linear heat T
generation rate is 15.02 kW/ft (Fn =2.32) for ENC TOPROD fuel. The maximum local metal-water reaction in this case is 5.60% after 380 seconds, and the total core metal-water reaction is less than 1%. The PCT location is at an elevation of 9.13 feet from the bottom of active core. For the exposure up to 15,000 MWD /MTM, the DCT is 20910F (Figure 2.33) including a 100F for UPI effect, occurring at 8.88 feet elevation relative to the bottom of the active core.
The local metal-water reaction is 4.68%, with a total metal-water reaction of less than 1%.
3 i
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XN-NF-83-38 i
Table 2.1 Prairie Island Units 1 and 2 System Data r.
Primary Heat Output, MWt 1650*
Primary Coolant Flow, lbm/hr 6.82 x 107 Primary Coolant Volume, ft3 10,309.**
Operating Pressure, psia 2,250.
Inlet Coolant Temperature, OF 530.
Reactor Vessel Volume, ft3 2364.
1 Pressurizer Volume, Total, ft3 1000.
Pressurizer Volume, Liquid, ft3 600.
Accumulator Volume, Total, ft3 (each of two) 2000.
Accumulator Volume, Liquid, ft3 1250.
Accumulator Trip Point Pressure, psia 714.7 Steam Generator Heat Transfer Area, ft2 48,925.***
Steam Generator Secondary Flow, lbm/hr 3.54 x 106 Steam Generator Secondary Pressure, psia 724.7 Reactor Coolant Pump Head, ft 277.
Reactor Coolant Pump Speed, rpm 1190.
2 Moment of Inertia, lbm-ft / rad 78,000.
Cold Leg Pipe, I.D., in 27.5 Hot Leg Pipe, I.D., in 29.0 I
Pump Suction Pipe, I.D., in 31.0 4.
- Primary Heat 0utput used in RELAP4-EM Model = 1.02 x 1650 = 1683 MWt.
- Includes total accumulator and pressurizer volumes.
Includes 5% SG tube plugging.
+
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9 XN-NF-83-38 Table 2.2 Fuel Design Parameters Parameter ENC Standard TOPROD l
Cladding, 0.D., in.
0.426 0.417 Cladding, I.D., in.
0.364 0.358 Cladding Thickness, in.
0.031 0.0295 Pellet 0.D., in.
0.3565 0.3505 Diametral Gap, in.
0.0075 0.0075 Pellet Density, % TO 94.0 94.0 Active Fuel Length, in.
144.0 144.0 Enriched UO, in.
144.0 132.0 2
6.0 Upper Blanket, in.
6.0 Lower Blanket, in.
l Cell Water / Fuel Ratio 1.67 1.79 l
Rod Pitch 0.556 0.556 e
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10 XN-NF-83-38 Table 2.3 Prairie Island Units 1 and 2 TOPROD LOCA-ECCS Analysis Results, Event Times Event Time (sec.)
Start 0.00 Break Initiation 0.05 Safety Injection Signal 0.65 Accumulator Injection, Intact Loop 8.70 Accumulator Injection, Broken Loop 4.80 End-of-Bypass 21.14 i
Safety Injection Flow 25.60 Start of Reflood 36.89 Accumulator Empties, Intact Loop 44.09 Peak Clad Temperature Reached 196.00 i
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11 XN-NF-83-38
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TIME AFTER BREAK ( SEC )
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J TIME - SECONDS Figure 2.33 T00DEE2 Cladding Temperature vs Time 0.4 DECLG Break, 0-15,000 MWD /MTM n
4 PRAIRIE ISLAND, MIP, T00DEE2, 0.4 DECLG 5% SG PLUGGING L
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ll Figure 2.34 T00DEE2 Cladding Temperature vs Time
- li l
0.4 DECLG Break, E0L
p 45 XN-NF-83-38
3.0 CONCLUSION
For breaks up to and including the double-ended severance of a reactor coolant pipe, the Emergency Core Cooling System for both Prairie Island units will meet the Acceptance Criteria as presented in 10 CFR 50.46, with the 2.32 Fhad1.55Ffglimits.
The criteria are as follows:
(1) The calculated peak fuel element clad temperature does not exceed the 22000F limit.
(2) The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1 percent of the total amount of zircaloy in the reactor.
(3) The cladding temperature transient is terminated at a time when the core geometry is still amenable to cooling.
The hot fuel rod cladding oxioation limits of 17% are not exceeded during or after quenching.
(4) The core temperature is reduced and decay heat is removed for an extended period of time, as required by tha long-lived radioactivity remaining in the core, f
46 XN-NF-83-38
4.0 REFERENCES
1.
ECCS Large Break Spectrum Analysis for Prairie Island Unit 1 Using ENC WREM-IIA PWR Evaluation Model, XN-NF-78-46, November 1978.
2.
Prairie Island Unit 2 Nuclear Plant Cycle 5 Safety Report, XN-NF-79-67, August 1979.
3.
LOCA ECCS Analysis for Prairie Island Unit 1 and 2 with ENC TOPROD Fuel, XN-NF-80-49, November 12, 1980.
4.
Exxon Nuclear Company Evaluation Model EXEM/PWR ECCS Model Updates, t
Xn-NF-82-20(P), Revision 1, August 1982; Supplement 1, Marc11982; I
and Supplement 2, March 1982.
5.
RODEX2: Fuel Rod Thermal-Mechanical Response Evaluation Model, XN-NF-81-58(P), Revision 2, February 1983.
6.
U. S. Nuclear Regulatory Commission, " Safety Evaluation Report on interim ECCS Evaluation Model for Westinghouse Two-Loop Plants,"
Analysis Branch, Division of System Safety, Office of Nuclear Reactor Regulation, November 1977.
7.
Letter, L. O. Mayer to Director of Nuclear Reactor Regulation, February 24, 1978 (Docket No. 50-282 and 50-306).
8.
" Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors," 10 CFR 50.46 and Appendix K of 10 CFR 50; Federal Register, Volume 39, Number 3, January 4,1974.
9.
Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model, XN-75-41, July 1975, and Supplements and Revisions thereto.
- 10. GAPEXX: A Computer Program for Predicting Pellet-to-Cladding Heat Transfer Coefficients, XN-73-25, August 13, 1973.
11.
U.S. Nuclear Regulatory Commission Letter, T. A. Ippolito (NRC) to W. S. Nechodom (ENC), "SER for ENC RELAP4-EM Update," March 1979.
12.
U.S. Nuclear Regulatory Commission, " Minimum Containment Pressure Model for PWR ECCS Performance Evaluation," Branch Technical
{
Position CSB 6-1.
\\
47 XN-NF-83-38
- 14. Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model, XN-NF-82-0/(P), Revision 1, August 1982.
15.
G.
N. Lauben, "T00DEE2: A Two-Dimensional Time Dependent Fuel Element Thermal Analysis Program," NRC Report NUREG-75/057, May 1975.
- 16. Exxon Nuclear Company ECCS Evaluation of a 2-Loop Westinghouse PWR with Dry Contaimnent Using the ENC WREM-II ECCS Model - Large Break Example Problem, XN-NF-77-25(A), September 1978.
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XN-NF-83-38 Issue Date:.5/27/83 PRAIRIE ISLAND UNITS 1 AND 2 LIMITING BREAK LOCA-ECCS ANALYSIS USING EXEM/PWR Distribution FT Adams GJ Busselman JC Chandler RE Collingham GC Cooke LJ Federico SE Jensen WV Kayser
'MR K111 gore TR Lindquist LC O'Malley GA Sofer RB Stout-T Tahvili PJ Valentine Northern States Power (80)/LC O'Malley P
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