ML20090K481
| ML20090K481 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 05/22/1984 |
| From: | Ades M, Braun D, Fausz N SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML17320B071 | List: |
| References | |
| XN-NF-84-21-(NP, XN-NF-84-21-(NP)-R01, XN-NF-84-21-(NP)-R1, NUDOCS 8405240220 | |
| Download: ML20090K481 (66) | |
Text
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XN-NF-84-21 (NP?
I REVISION 1 DONALD C. COOK UNIT 2 CYCLE 6 5o/o STEAM GENERATOR TUBE PLUGGING LIMITING BREAK LOCA/ECCS ANALYSIS l
l MAY 1984 RICHLAND,WA 99352 ERON NUCLEAR COMPANY, INC.
[885!!8!!J888),fg
1 XN-NF-84-21 (NP )
Revision 1 Issue Date: 5/22/84 DONALD C. COOK UNIT 2 CYCLE 5 - 5% STEAM GENERATOR TUBE PLUGGING LIMITING BREAK LOCA/ECCS ANALYSIS Prepared by:
,c M
$ % fd 44 MJ1 des, 00/Braun, NF Faust (/JE fajicek Licensing & Safety Engineering Approve:
W4 W. V. Kafser, Manager PWR Safety Analysis
[h[f/
Concur:
J. C. Chandler, Lead Enginee~r Reload Fuel Licensing 1
Approve:
2201A3/9 l
R. B. Stout, Manager Licensing & Safety Engineering I
l Concur:
, [
27 d> M" h l
J. N y. organ,~ Manager Prop 6sals & Customer Services Engineering Approve:
eD.2 AfA Y F V G. A. S K, Manager Fuel Engineering & Technical Services j
gf E(ON NUCLEAR COMPANY,Inc.
l l
0 NUCLEAR htGULATORY COMMISSION OISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY
(
This technical report was dertwed 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-bution to facilitate safety analyses by licensees of the USNRC which utilize Exxon Nucleer fabricated reload fuel or other technical services 5
provided by Exxon Nuclear for licht water power reactors and it is fra and correct a the best of Exxon Nucleer's knowledge, informacon, and belief. The information contained herein may be used by the USNRC in its review of this report, and by tiansees or applicants before the USNRC which are cussomers of Exxon Nudeer in their demonstration of comoliance with the USN RC's regulations.
Without derogating from the foregoing neither Exxon Nuclear nor any oorson acting nn its behalf:
A.
Mekas any warranty, express or implied, with respect to the acoaracy, comph or usefulness of the infor-motion contained in this document, or that the use of any informatiort apparatus, method, or process disclosed in this document will not infringe privatefy owned rights; or I
l 8.
Assumes any liabilitus with respect to the use of, or for l
darrages reeJ! ting from the use of, any information, ap-paratus, method, or process disclosed in this document.
l XN-NF-F00, 766 I
i XN-NF-84-21 ( NP )
Revision 1 l
h TABLE OF CONTENTS h
Sect ion Page
1.0 INTRODUCTION
I h
2.0
SUMMARY
2 3.0 LIMITING BREAK LOCA ANALYSIS.......................
4
,3.1 LOCA ANALYSIS MODEL...........................
4 3.2 RESULTS.......................................
6 4
4.0 CONCLUSION
S........................................
54
5.0 REFERENCES
55 l
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r XN-NF-84-21' NP )
Revision 1 1
l LIST OF TABLES 1
Table Page 2.1 D.C. Cook Unit 2 LOCA/ECCS Analysis Sunnary........
3 3.1 Donald C. Cook Unit 2 System Input Parameters......
7 3.2 1.0 DECLG Break Analys is Parameters................
8 3.3 D.C. Cook Unit 2 1.0 DECLG Break Event Times.......
9 3.4 1.0 DECLG Break Fuel Response Results for Cycle 5............................................
10
(
i ii XN-NF-84-21 (NP)
Revision 1
(
LIST OF FIGURES 4
i l
Figure Page l
l t
3.1 RELAP4/EM Blowdown System Nodalization s
for D.C. Cook Unit 2..............................
11 f
3.2 Downcomer Flow Rate During Blowdown Period, 1.0 DECLG Break...................................
12
)
3.3 Upper Plenum Pressure during Blowdown I-Period, 1.0 DECLG Break...........................
13 3.4 Average Core Inlet Flow during Blowdown Period, 1.0 DECLG Break...........................
14 3.5 Average Core Outlet Flow during Blowdown Period, 1.0 DECLG Break...........................
15 3.6 Total Break Flow during Blowdown Period, 1.0 DECLG Break...................................
16 3.7 Break Flow Enthalpy during Blowdown, 1.0 DECLG Break...................................
17 l
3.8 Flow from Intact Loop Accumulator during Blowdown Period, 1.0 DECLG Break..................
18 l
3.9 Flow from Broken Loop Accumulator during i
Blowdown Period, 1.0 DECLG Break..................
19 3.10 Pressurizer Surge Line Flow during Blowdown
. Per iod, 1.0 DECLG Bre ak...........................
20 j
3.11 Heat Transfer Coefficient during Blowdown Period at PCT Node, 1.0 DECLG Break, 2.0 MWD /kg Case...................................
21 t
i-3.12 Clad Surface Temperature during Blowdown Period at PCT Node, 1.0 DECLG Break, 2.0 MWD /kg Case...................................
22 3.13 Depth of Metal-Water Reaction during Blowdown Period at PCT Node, 1.0 DECLG Break, 2.0 MWD /kg Case...................................
23 E
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iv XN-NF-84-21(NP )
Revision 1 LIST OF FIGURES (Cont.)
Figure Page 3.14 Average Fuel Temperature during Blowdown Period at PCT Location, 1.0 DECLG Break, 2.0 MWD /kg Case....................................
24 s
3.15 Hot Assembly Inlet Flow during Blowdown Period, 1.0 DECLG Break, 2.0 MWD /kg Case...........
25 3.16 Hot Assembly Outlet Flow during Blowdown Period, 1.0 DECLG Break, 2.0 MWD /kg Case...........
26
(
3.17 Heat Transfer Coefficient during Blowdown Period at PCT Node, 1.0 DECLG Break, 10.0 MWD /kg Case...................................
27 3.18 Clad Surface Temperature during Blowdown Period at PCT Node, 1.0 DECLG Break, 10.0 MWD /kg Case...................................
28 3.19 Depth of Metal-Water Reaction during Blowdown Period at PCT Node, 1.0 DECLG Break, l
10.0 MWD /kg Case...................................
29 3.20 Average Fuel Temperature during Blowdown Period at PCT Location, 1.0 DECLG Break, 10.0 MWD /kg Case...................................
30 3.21 Hot Assembly Inlet Flow during Blowdown Period, l
1.0 DECLG Break, 10.0 MWD /kg Case..................
31 i
3.22 Hot Assembly Outlet Flow during Blowdown Period, l
1.0 DECLG Break, 10.0 MWD /kg Case..................
32 3.23 Heat Transfer Coefficient during Blowdown Period at PCT Node, 1.0 DECLG Break, j
47.0 MWD /kg Case...................................
33 l
3.24 Clad Surface Temperature during Blowdown Period at PCT Node, 1.0 DECLG Break, 47.0 MWD /kg Case...................................
34 l
3.25 Depth of Metal-Water Reaction during Blowdown Period at PCT Node,1.0 DECLG Break, l
47.0 MWD /kg Case...................................
35 l
l l
l l
" XN-NF-84-21(NP) v Revision 1 LIST OF FIGURES (Cont.)
l Figure Page f
3.26 Average Fuel Temperature during Blowdown Period at PCT Location,1.0 DECLG Break, 47.0 MWD /kg Case..................................
36 3.27 Hot Assembly Inlet Flow during Blowdown Period, 1.0 DECLG Break, 47.0 MWD /kg Case.................
37 3.28 Hot Assembly Outlet Flow during Blowdown Period, 1.0 DECLG Break, 47.0 MWD /kg Case.........
38
)
3.29 Accumulator Flow during Refill and Reflood Periods, Broken Loop, 1.0 DECLG Break.............
39 3.30 Accumulator Flow during Refill and Reflood Periods, Intact Loop,.1.0 DECLG Break.............
40 3.31 HPSI & LPSI Flow during Refill and Reflood Periods, Broken Loop, 1.0 DECLG Break.............
41 3.32 HPSI & LPSI Flow during Refill and Reflood Periods, Intact Loop, 1.0 DECLG Break.............
42 3.33 Containment Back Pressure, 1.0 DECLG Break........
43 3.34 Nonnalized Power,1.0 DECLG Break, 44 2.0 MWD /kg. Case...................................
3.35 Normalized Power,1.0 DECLG Break, 10.0 MWD /kg Case..................................
45 3.36 Normalized Power,1.0 DECLG Break, 46 47.0 MWD /kg Case..................................
3.37 Reflood Core Mixture Level,1.0 DECLG Break, 85% ENC Core..............................
47 3.38 Reflood Downcomer Mixture Level, 1.0 DECLG Break...................................
48 49 3.39 Reflood Upper Plenum Pressure,1.0 DECLG Break....
50 3.40 Core Floodi'ng Rate, 1.0 DECLG Break...............
T -
vi XN-NF-84-21(NP )
Revision 1 LIST OF FIGURES (Cont.)
Figure Page 3.41 T00DEE2 Cladding Temperature versus Time, 1.0 DECLG Break, 2.0 MWD /kg Case..................
51 1
3.42 T000EE2 Cladding Temperature versus Time, J
1.0 DECLG Break, 10.0 MWD /kg Case.................
52 3.43 T000EE2 Cladding Temperature versus Time, l.0 DECLG Break, 47.0 MWD /kg Case.................
53 1
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1 XN-NF-84-21 (NP )
Revision 1 r
1.0 INTRODUCTION
Large break LOCA/ECCS analyses were performed in 2982(1,2) to support operation of the D.C. Cook Unit 2 reactor at 3425 MWt with ENC fuel, Reference s
1 presented analytical results for a spectrum of postulated large break LOCAs.
The limiting break was identified as the 1.0 DECLG break.
Reference 2 l
presented results for the previously identified limiting break using the EXEM/PWR(3) ECCS models, except GAPEX was used as the fuel performance model in place of RODEX2. The RODEX2 code was not approved by the NRC for use in ECCS analyses in 1982.
The analysis therefore used the GAPEX(4) code which was approved by the NRC to calculate fuel properties at the initialization of the LOCA calculation.
The Reference 2 report documented the results of calculations with one and two LPSI pumps operating.
At equivalent core peaking limits, higher peak cladding temperatures (PCTs) were calculated in the LOCA analysis when two LPSI pumps were assumed operating. The Reference i
2 analysis with two LPSI pumps operating was performed for Cycle 4 operation of D.C. Cook Unit 2.
This report documents the results of a LOCA/ECCS analysis to support operation of the D.C. Cook Unit 2 reactor for Cycle 5 at a thermal power rating of 3425 MWt, with up to 5% of the steam generator tubes plugged, with two LPSI pumps operating, and for ENC fuel exposed up to a peak rod average burnup of 47 MWD /kg.
The calculations were performed using the EXEM/PWR LOCA/ECCS models, including fuel properties calculated at the start of the LOCA i
l transient with ENC's generically approved RODEX2 code.(5) l l
k F
2 XN-NF-84-21 (NP )
Revision 1 f
2.0 SlNMARY LOCA/ECCS calculations were performed to determine core peaking limits which permit operation of the D.C. Cook Unit 2 reactor within guidelines specified by 10 CFR 50.46 and Appendix K.(6)
The calculations assumed operation:
1)
At a thermal power of 3425 MWt; 2)
With 5% average steam generator tube plugging; and 3)
With the Cycle 5 core configuration (85% ENC fuel).
The calculations were performed for the previously identified limiting break, the 1.0 DECLG break, with full ECCS flow.
The results of the analysis are summarized in Table 2.1.
The analysis supports operation of the D.C. Cook Unit 2 reactor for Cycle 5 at a total peak limit (Fg ) of 2.04 and a corresponding FfH T
limit of 1.415.
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3 XN-NF-84-21 (NP)
Revision 1 Table 2.1 D.C. Cook Unit 2 LOCA/ECCS Analysis Sumary Results for the Cycle 5 Core Configuration (85% ENC Fuel)
Peak Rod Average Burnup (MWD /kg) 2.0
, 10.0 47.0 Fh 2.04 2.04 2.04 FfH 1.415 1.415 1.415 i
Peak Cladding Temperature (OF) 2198 2190 2096 Maximum 1ocal Zr-H O Reaction (%)
7.4 7.3 5.7 2
Total Zr-H O Reaction
< 1.0
< 1.0
< 1.0 2
l e
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3 gyw- - -
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4 XN-NF-84-21 (NP )
Revision 1 3.0 LIMITING BREAK LOCA ANALYSIS This report supplements previous LOCA/ECCS analyses performed and I
documented for D.C. Cook Unit 2.
A spectrum of LOCA breaks was performed and reported in XN-NF-82-35.(1) The limiting LOCA break was determined to be the l
large double-ended guillotine break of the cold leg or reactor vessel inlet pipe with a discharge coefficient of 1.0 (1.0 DECLG). Reference 2 established I
that for D.C. Cook Unit 2 it is more limiting in the LOCA analysis to assume no failure of a LPSI pump.
The analysis performed and reported herein considers:
1)
That 5% of the steam generator tubes are plugged; 2)
That 85% of the Cycle 5 core is composed of ENC fuel; 3)
That both LPSI pumps are operational; and 4)
That ENC fuel may be exposed to a peak average burnup of 47 MWD /kg.
3.1 LOCA ANALYSIS MODEL The Exxon Nucleat Company EXEM/PWR-ECCS evaluation model was used to perform the analyses required.
This model(3) consists of the following computer codes: RODEX2(5) code for initial stored energy; RELAP4-EM(7) for the system blowdown and hot channel blowdown calculations; ICECON(8) for the computation of the ice condenser containment backpressure; REFLEX (3,9) for computation of system reflood; and T00DEE2(3,10,11) for the calculation of final fuel rod heatup.
The Donald C.
Cook Unit 2 nuclear power plant is a 4-loop Westinghouse pressurized water reactor with ice condenser containment. The reactor coolant system is nodalized into control volumes representing reasonably homogeneous regions, interconnected by flow-paths or " junctions".
5 XN-NF-84-21(NP)
Revision 1 reasonably homogeneous regions, interconnected by flow-paths or " junctions".
The system nodalization is depicted in Figure 3.1.
The unbroken loops were assumed symmetrical and modeled as one intact loop with appropriately scaled input. Pump performance curves characteristic of a Westinghouse series 93A pump were used in the analysis. The transient behavior was determined from the governing conservation equations for mass, energy, and momentum. Energy transport, flow rates, and heat transfer were determined from appropriate correlations.
The Cycle 4 LOCA analysis (2) assumed that 1% of the steam generator tubes were plugged. In the current analysis, the plant was modeled assuming asymmetric steam generator tube plugging: 3.33% of the tubes plugged in the intact loops, and 10.0% of the tubes plugged in the broken loop. The larger plugging in the broken loop results in higher PCTs. The primary coolant flow at full power was reduced by 1.1% from the current measured flow at the plant to account for the assumed average 5% steam generator plugging. Additionally, the core model assumed that the core is 85% ENC fuel, whereas the previous analysis assumed the Cycle 4 core configuration. ENC fuel has a smaller rod diameter than the Westinghouse fuel it replaces.
To offset the impact of increased flow area on the LOCA analysis results, the core power was reduced from 3425 MWt to 3411 MWt.
System input parameters are given in Table 3.1.
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. Chopped cosine axial power
6 XN-NF-84-21 ( NP )
Revision 1 profiles are assumed with the maximum axial peaking factor used in the analysis given in Table 3.2.
The analysis of the loss-of-coolant accident is performed at 102 percent of rated power. The core power and other parameters used in the analyses are given in Table 3.1.
3.2 RESULTS Table 3.3 presents the timing and sequence of events as determined for the large break guillotine configuration with a discharge coefficient of 1.0 for full ECCS operation. Table 3.4 presents the results of the exposure analysis for Cycle 5 composed of 85% ENC fuel.
Results' of the analyses are given in Figures 3.2 to 3.43. Figures 3.2 to 3.10 provide plots of key system blowdown parameters versus times.
Figures 3.11 to 3.28 provide plots at key core responses during the blowdown period. Figures 3.29 to 3.32 provide the ECCS flows in the broken and intact loopduringthebefillperiod. Figure 3.33 presents the containment pressure during the LOCA. Figures 3.34 to 3.36 present the normalized power during the LOCA for the three exposure cases analyzed.
Figures 3.37 to 3.40 provide results from the reflood portion of the transient for the case in which 85% of the core is ENC fuel. Finally, Figures 3.41 to 3.43 provide the response of the fuel during the refill and reflood periods of the LOCA transient for the fuel burnup cases investigated.
7 XN-NF-84-21 (NP)
Revision 1 Table 3.1 Donald C. Cook Unit 2 System Input Parameters Thermal Power, MWt*
3425 Core, MWt 3411 Pump,"MWt 14 1
Primary Coolant Flow, Mlbm/hr 143.1 Primary Coolant Volume, ft3 31,768 Operating Pressure, psia 2250 Inlet Coolant Temperature, OF 542 Reactor Vessel Volume, ft3 4945 Pressurizer Volume, Total, ft3 1800 Pressurizer Volume, Liquid, ft3 1080 Accumulator Volume, Total, ft3 (each of four) 1350 Accumulator Volume, Liquid, ft3 (each of four) 950 Accumulator Pressure, psia 636 Steam Generator Heat Transfer Area, ft2_
SG1, SG2, SG3, SG4 11,588,3(12,446) 6 Steam Generator Secondary Flow, lbm/hr -
3.505 x 10,
6 SG1, SG2,. SG3, SG4 3(3.764 x 10 )
l Steam Generator Secondary Pressure, psia 799 Reactor Coolant Pump Head, ft 277 Reactor Coolant Pump Speed, rpm 1189 Moment of Inertia, lbm-ft2 82,000 Cold Leg Pipe, I.D. in.
27.5 Hot Leg Pipe, I.D. in.
29.0 Pump Suction Pipe, I.D. in.
31.0 Fuel Assembly Rod Diameter, in.
0.360 Fuel Assembly Rod Pitch, in.
0.496 Fuel Assembly Pitch, in.
8.466 Fueled (Core) Height, in.
144.0 2
Fuel Heat Transfer Area, ft **
57,327 2
Fuel Total Flow Area, Bare Rod, ft **
53.703 Refueling Water Storage Tank Temperature, OF 80 Accumulator Water Temperature, OF 120
- Primary. Heat Output used in RELAP4-EM Model = 1.02 x 3425 = 3493.5 MWt
- ENC Fuel Parameters.
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8 XN-NF-84-21(NP)
Revision 1 Table 3.2 1.0 DECLG Break Analysis Parameters Peak Rod Average Burnup (MWD /kg) 2.0 10.0 47.0 Total Core Power (MWt)*
3411 3411 3411 Total Peaking (F )
2.04 2.04 2.04 Fraction Energy Deposited in Fuel
. Fully Moderated Core 0.974 0.974 0.974
. Voided Core 0.954 0.954 0.954 Cycle 5 (85% ENC Fuel)
Peaking
. Axial x Engineering 1.442 1.442 1.442
.EnthalpyRise(FIH) 1.415 1.415 1.415
- 2% power uncertainty is added to this value in the LOCA analysis.
9 XN-NF-84-21 (NP',
Revision 1 Table 3.3 D. C. Cook Unit 21.0 DECLG Break Event Times j
Event Time (sec.)
Start 0.00 Break Initiation 0.05 Safety Injection Signal 0.65 Accumulator Injection
)
Broken Loop 3.2 Intact Loop 15.5 End of Bypass 24.31 Safety Pump Injection 25.65 Start of Reflood 40.48 Accumulator Empty Broken Loop 44.2 i
Intact Loop 52.9 l
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l 10 XN-NF-84-21 (NP)
Revision 1 i 's Table 3.4 1.0 DECLG Break Fuel Response Results for Cycle 5 l
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Peak Rod Average Burnup (NWD/kg) 2.0 -
10.0 47.0 l
Initial Peak Fuel Average Tssperature (OF) 2151 2060 1629 l
Hot Rod Burst L
. Time _(sec) 60.9 61.7 67.9 s
. Elevation (ft) 6.50 6.50 7.00
. Channel Blockage Fraction
.' 24
.27
.47 Peak Clad Temperature w.
. Time (sec) 227 227 241
.;. Elevation (ft) 8.63 8.63 8.88
. Temperature (OF) 2198 2190 2096 Zr-Steam Reaction
. Local Maximum Elevation (ft) 8.63 8.63 8.88
. Local Maximum (%)*
7.4 7.3 5.7
. Core Maximum
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54 XN-NF-84-21(NP)
Revision 1
4.0 CONCLUSION
S For breaks up to and including the double-ended severance of a reactor coolant pipe, the Donald C. Cook Unit 2 Emergency Core Cooling System will meet the Acceptance Criteria as presented in 10 CFR 50.46 for operation with ENC 17x17 fuel operating in accordance with the LHGR limits noted in Table 2.1.
That is:
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 oxidation 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 the long-lived radio-activity remaining in the core.
1
55 XN-NF-84-21(NP)
Revision 1
5.0 REFERENCES
(1) XN-NF-82-35, " Donald C. Cook Unit 2 LOCA ECCS Analysis Using EXEM/PWR Large Break Results," Exxon Nuclear Company, Inc., Rich-land, WA 99352, April 1982.
(2) XN-NF-82-35, Supplement 1, " Donald C. Cook Unit 2 Cycle 4 Limiting Break LOCA-ECCS Analysis Using EXEM/PWR," Exxon Nuclear Company, Inc., Richland, WA 99352, November 1982.
(3) XN-NF-82-20(P), Rev.1, August 1982; Supplement 1, March 1982; and Supplement 2, March 1982, " Exxon Nuclear Company Evaluation Model EXEM/PWR ECCS Model Updates," Exxon Nuclear Company, Inc., Rich-land, WA 99352.
(4) XN-73-25, "GAPEXX: A Computer Program for Predicting Pellet-to-Cladding Heat Transfer Coefficients," Exxon Nuclear Company, Inc.,
Richland, WA, August 13, 1973.
(5) XN-NF-81-58(P), Rev. 2, "RODEX2: Fuel Rod Thermal-Mechanical Re-sponse Evaluation Model," Exxon Nuclear Company, Inc., Richland, WA 99352, February 1983.
(6) " 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.
(7)
U.S. Nuclear Regulatory Commission letter, T.A. Ippolito (NRC) to W.S. Nechodom (ENC), "SER for ENC RELAP4-EM Update," March 1979.
(8) XN-CC-39, Rev. 1, "ICECON: A Computer Program Used to Calculate Containment Backpressure for LOCA Analysis (Including Ice Condenser Plants)," Exxon Nuclear Company, Inc., Richland, WA 99352, November 1977.
(9)
XN-NF-78-30( A), " Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model Update ENC WREM-IIA," Exxon Nuclear Company, Inc.,
Richland, WA 99352, May 1979.
(10)
XN-NF-82-07(A), Rev.
1,
" Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model," Exxon Nuclear Company, Inc., Richland, WA 99352, March 1982.
(11)
G.N. Lauben, NRC Report NUREG-75/057, "T00DEE2: A Two-Dimensional Time Dependent Fuel Element Thermal Analysis Program," May 1975.
56 XN-NF-64-21(NP)
Revision 1 (12)
D.C. Cook Unit 2 Technical Specification, Appendix "A" to License No.
DPR-74, Amendment No. 48.
(13)
XN-NF-82-32(P), Revision 2, " Plant Transient Analysis for the Donald C. Cook Unit 2 Reactor at 3425 MWt: Operation with 5% Steam Generator Tube Plugging," Exxon Nuclear Company, Inc., Richland, WA 99352, February 1984.
XN-NF-84-21(NP)
Revision 1 Issue Date: 5/22/84 DONALD C. COOK UNIT 2 CYCLE 5 5% STEAM GENERATOR TUBE PLUGGING LIMITING BREAK LOCA/ECCS ANALYSIS Distribution M. J. Ades D. J. Braun J. C. Chandler R. A. Copeland i
N. F. Fausz S. E. Jensen W. V. Kayser J. E. Krajicek G. F. Owsley H. G. Shaw G. A. Sofer R. B. Stout T. Tahvili AEP/H.G. Shaw (10)
USNRC/J.C. Chandler (41)
Document Control (5)
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