ML17319B666
| ML17319B666 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 11/18/1982 |
| From: | JENSEN S E, KAYSER W V, TANVILI T SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML17319B667 | List: |
| References | |
| XN-NF-82-35-S01, XN-NF-82-35-S1, NUDOCS 8211290375 | |
| Download: ML17319B666 (61) | |
Text
8>>>~9p375 ski>zq.PDR-ADOCK pgppppi~PDR XN-NF-82-35 Supplement 1 Issue Oate: 11/18/82 OONALD C.COOK UNIT 2 CYCLE 4 LIMITING BREAK LOCA-ECCS ANALYSIS USING EXEM/PWR Prepared by: avii NSSS Systems Analysis (ECCS)Reviewed by: nsen, anager NSSS Systems Analysis (ECCS)Concur: J/p Fz, W.V.Kayser, anager Fuel Response Analysis Concur: s ey, a ager Reload Fue Li ensing Approve: tout, anager Licensing L Safe Engineering Approve: o er, Manager Fuel Engineering
&Technical Services:mb E)5(ON NUCLEAR COMPANY, Inc.
NUCLEAR REGULATORY COMMISSION DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULI.Y This technical report was rlerived 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 Nudear-fabricated reload fuel or other technical services provided by Exxon Nuclear for liaht 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 demonstradon of comoliance with the USNRC's regulations.
Without demgadng from the foregoing, neither Exxon Nuclear nor any person acting nn its behalf: A.Makes any warranty, express or implied, with respect to the accuracy, completeness, or usefulness of the infor-mation contained in this document, 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 for dan ages resulting from the use of, any information, ap-paratus, method, or process disclosed in this document.XN-NF-FOO, 766 TABLE OF CONTENTS XN-NF-82-35 Supplement 1 SECTION PAGE
1.0 INTRODUCTION
AND
SUMMARY
2.0 LIMITING
BREAK LOCA ANALYSES-2.1 LOCA ANALYSIS MODEL 2.2 RESULTS
3.0 CONCLUSION
S
4.0 REFERENCES
.51 52 XN-NF-82-35 Supplement 1 LIST OF TABLES TABLE PAGE 1.1 Donald C.Cook Unit 2, Cycle 4, DECLG (CD=1.0)Break Analysis for Full ECCS Flow, and One LPSI Pump Operation 2.1 Donald C.Cook Unit 2 System Input Parameters
..........
2.2 Donald
C.Cook Unit 2, Cycle 4, DECLG (CD=1.0)Break Events for Full ECCS Flow, and One LPSI Pump Flow LIST OF FIGURES XN-NF-82-35 Supplement 1 FIGURE PAGE 2.1 System Blowdown Nodalization for the Donald C~Cook Unit 2 PWR.2.2 Axial Peaking Factor versus Rod Length 1.0 DECLG Break with Full ECCS Flow.2.3 Axial Peaking Factor versus Rod Length 1.0 OECLG Break with Single Failure ECCS Flow 2.4 Upper Plenum Pressure 1~0 OECLG Break (Single Failure and Full ECCS Flow)2.5 Pressurizer Pressure 1.0 DECLG Break (Single Failure and Full ECCS Flow)2.6 Total Break Flow 1.0 DECLG Break (Single Failure and Full ECCS Flow)10 12 13 14 15 2'Average Core Inlet Flow 1.0 OECLG Break (Single Failure).2.8 Average Core Outlet Flow 1.0 DECLG Break (Single Failure)16 17 2.9 Downcomer Flow Rate 1.0 OECLG Break (Single Failure E Full ECCS Flow)............
18 2.10 Pressurizer Surge Line Flow 1.0 DECLG Break (Single Failure 5 Full ECCS Flow)............
19 2.11 Flow From Intact Loop Accumulators
1.0 DECLG
Break (Single Failure&Full ECCS Flow)............
20 2.12 Flow From Broken Loop Accumulators
1.0 DECLG
Break (Single Failure 8 Full ECCS Flow)............
21 2.13 Hot Channel Average Fuel Temperature
1.0 OECLG
Break (Single Failure)2.14 Clad Surface Temperature
1.0 OECLG
Break (Single Failure)2'5 Depth of Metal-Mater Reaction 1~0 DECLG Break (Single Failure)22 23~~~~~~~~~~~~24 FIGURE 1V I XN-NF-82-35 Supplement 1 PAGE 2.16 Hot Channel Heat Transfer Coefficient
1.0 OECLG
Break (Single Failure)..~..~...~...................
25 2.17 Hot Assembly Inlet Flow 1.0 DECLG Break (Single Failure)2.18 Hot Assembly Outlet Flow 1.0 DECLG Break (Single Failure)2.19 ICECON Containment Back Pressure 1.0 OECLG Break (Single Failure)26~~~~~~~~~~~~~~27 28 2.20 Normalized Power 1.0 OECLG Break (Single Failure and Full ECCS Flow)......'...
2g 2.21 Core Flooding Rate 1.0 DECLG Break (Single Failure)2.22 Ref lood Downcomer Mixture Level 1.0 OECLG Break (Single Failure)30 31 2'3 Reflood Core Mixture Level 1.0 DECLG Break (Single Failure).........32 2.24 Reflood Upper Plenum Pressure 1.0 DECLG Break (Single Failure)............................
33 2'5 Reflood Core Saturation Temperature
1.0 OECLG
Break (Single Failure)2.26 TOODEE2 Cladding Temperature vs.Time 1.0 DECLG Break (Single Failure)..2.27 Average Core Inlet Flow 1.0 OECLG Break (Full ECCS Flow)2.28 Average Core Outlet Flow 1.0 DECLG Break (Full ECCS Flow)2.29 Hot Channel Average Fuel Temperature
1.0 DECLG
Break (Full ECCS Flow)..2.30 Clad Surface Temperature
1.0 DECLG
Break (Full ECCS Flow).1 Depth of Metal-Water Reaction 1.0 OECLG Break (Full ECCS Flow)..2.32 Hot Channel Heat Transfer Coefficient
1.0 DECLG
Break (Full ECCS Flow).34 35 36 37 38 39 40 41 XN-NF-82-35 Supplement 1 FIGURE PAGE 2.33 Hot Assembly Inlet Flow 1.0 DECLG Break (Full ECCS Flow)2.34 Hot Assembly Outlet Flow'.0 DECLG Break (Full ECCS Flow)2.35 ICECON Containment Back Pressure 1.0 DECLG Break (Full ECCS Flow)2.36 Core Flooding Rate 1.0 DECLG Break (Full ECCS Flow)2.37 Reflood Downcomer Mixture Level 1.0 DECLG Break (Full ECCS Flow)2.38 Ref lood Core Mixture Level 1.0 DECLG Break (Full ECCS Flow)2.39 Ref lood Upper Plenum Pressure 1.0 DECLG Break (Full ECCS Flow)2.40 Ref lood Core Saturation Temperature
1.0 DECLG
Break (Full ECCS Flow)2.41 TOODEE2 Cladding Temperature vs.Time 1.0 DECLG Break (Full ECCS Flow)42 43 44~---~--45 46 47 48 49 50' XN-NF-82-35 Supplement 1
1.0 INTRODUCTION
AND
SUMMARY
This document presents additional analytical results for a postulated large break loss-of-coolant accident (LOCA), performed for the Donald C.Cook Unit 2 nuclear power plant, operating at 3425 MWt, and fueled by Exxon Nuclear Company (ENC).Calculations.were performed qsing the EXEM/PWR ECCS Evaluation Model)with the RODEX2 stored energy model replaced by the (I)GAPEX model which has been previously approved by the United States (2)Nuclear Regulatory Commission (NRC).The analyses are applicable to Cycle 4 operation of D.C.Cook Unit 2 and include both the worst single failure assumption of loss of one low pressure safety injection (LPSI)pump and the no single failure assumption of full ECCS operation.
The results of the analyses show that within the limits established, the criteria specified by 10 CFR 50.46(3)are satisfied.
The break spectrum calculations for the large break LOCAs were pre-viously reported and the limiting break from the spectrum analysis was shown to be the large double-ended cold leg guillotine (DECLG)break with a discharge coefficient of 1.0.The limiting break was recalculated with the EXEM/PWR model and the GAPEX stored energy model.For this break, with the assumption that one of the LPSI pumps has failed and with a total power peaking factor (FqT)of 2.10, the Peak Cladding Temperature (PCT)is 2091oF,~occurring at 282 seconds into the accident at a location 9.13 feet from the bottom of the active core.With no failure of LPSI pumps assumed the 1.0 DECLG break LOCA yielded a PCT of 2198oF, which occurred 341 seconds into the accident at 9.63 feet from the bottom of the active core.The maximum linear heat generation rate for the no failure case with full ECCS flow is 11.66 kW/ft, including 1.02 factor for power uncertainties, which corresponds to a t XN-NF-82-35 Supplement 1 total peaking factor (Fq)of 2.04.Table 1.1 shows the calculated peak T cladding temperatures and metal-water reaction results for both cases.In all cases, the emergency core cooling system is shown to meet the Acceptance Criteria as presented in 10 CFR 50.46.For the limiting large break LOCA bounding.exposure conditions were assumed for the cycle.These bounding conditions included the highest initial fuel temperatures and greatest fission gas release, for Cycle 4 operation with ENC fuel.The maximum exposure used for the maximum power fuel rod was 22.11 GWD/MTM.
Table 1.1 Donald C.Cook Unit 2, Cycle 4, DECLG (CD=1.0)Break Analysis for Full ECCS Flow, and One LPSI Pump Operation.
Anal sis Results Peak Clad Temperature (PCT)oF Time of PCT, sec.Peak Clad Temperature location, ft.Local 2r/H20 Reaction (max.), X*Local Zr/H20 Location, ft.from bottom Total H2 Generation, X of total 2r Reacted Hot Rod Burst Time, sec.Hot Rod Burst Location, ft.Single Failure of~051 2091.282 9'3 5.91 9.13<1.0 59.57 6.75 Full ECCS Flow 2198.341.9'3 7.62 9.38<1.0 62.97 7.00 Calculation License Core Power, MWt Power Used for Analysis, MWt Peak Linear Power for Analysis, kW/ft**Total Peaking Factor, FqT Enthalpy Rise, Nuclear, FN<H Peak Rod Burnup Maximum (GWD/MTM)3425 3493.5 12.0 2.10 1.55 22.11 3425 3493.5 11.66 2'4 1.55 22.11*Computer value at 400 seconds.**Including 1.02 factor for power uncertainties XN-NF-82-35 Supplement 1 2.0 LIMITING BREAK LOCA ANALYSIS This report supplements previous LOCA-ECCS analyses performed and documented for D.C.Cook Unit 2.An example application of the EXEM/PWR model was reported in XN-NF-82-20(P)
Supplement 2(1).A spectrum of LOCA breaks was performed and reported in XN-NF-82-35'(4).
The limiting LOCA break was determined to be the large double-ended guillotine break of the cold leg or reactor vessel inlet pipe with a discharge coefficient of 1.0 (1.0 DECLG).The analyses performed and reported herein consider: (1)A revised stored energy model (GAPEX)in place of the previously applied RODEX2 model.The RODEX2 model as applied in the documented analyses(4) requires revision for NRC approval.Until a satisfactory resolution of RODEX2 is achieved, ENC will perform EXEM/PWR calculations using the currently approved GAPEX model for a limited exposure of one cycle.(2)Both the cases of a single failure of a low pressure safety ,.(LPSI)pump as determined by the NSSS vendor, and the full case.injection ECCS flow (3)Updates to the latest D.C.Cook Unit 2 application to reflect all model revisions as documented in XN-NF-82-20(P)
Revision 1(1).2.1 LOCA ANALYSIS MODEL The Exxon Nuclear Company EXEM/PWR-ECCS evaluation model was used to perform the analyses required.This model'(>)consist of the following computer codes: GAPEX(2)code for initial stored energy;RELAP4-EM(5) for the system blowdown and hot channel blowdown calculations; ICECON(6)for the computation of the ice condenser containment backpressure; REFLEX(1~7) for computation of system ref lood;and TOODEE2(1~Big), for the calculation of final fuel rod heatup.
~~XN-NF-82-35 Supplement 1 The Donald C.Cook Unit 2 nuclear power plant is a 4-loop West-inghouse 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" as described in XN-NF-82-20(P), Supplement 2(1).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.Pump performance curves characteristic of a Westinghouse series 93A pump were used in the analysis.The transient behavior was determined from the governing conser-vation equations for mass, energy, and momentum.Energy transport, flow rates, and heat transfer were determined from appropriate correlations'ystem 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 axial power profiles used for the analyses are shown in Figures 2.2 and 2.3, with a maximum axial peaking factor of 1.316 for the full ECCS case, and 1.355 for the single failure of one of the LPSI pumps.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 2.1.Both ECCS calculations were performed with input which bounds the fuel history for Cycle 4.The most limiting fuel conditions from beginning-of-life to end-of-cycle exposure conditions (22,110 MWD/MTM peak rod burnup)were determined and used in the calculation.
Internal rod pressure and decay power are highest at EOC for the hot rod while stored energy was calculated to be highest at beginning-of-life This combination of highest stored energy, highest rod pressure, and greatest decay power were used to bound Cycle 4 XN-NF-82-35 Supplement 1 operations The single failure calculation assumed an all-ENC fueled core while the full ECCS flow case used the D.C.Cook Unit 2 Cycle 4 core configuration.
These calculations will be redone for future cycles when the RODEX2 code for the stored energy model is approved by the NRC.2.2 RESULTS Table 2.2 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, and also for single failure of one LPSI pump.Prior to pumped safety injection initiation, the transient timing is the same for both cases.After safety injection begins, calculated results differ due to the increased ECCS mass flow for the full ECCS case relative to the single failure case.The principal effects of high ECCS flow on the evaluation model LOCA-ECCS calculation for D.C.Cook Unit 2 are a decreased containment pressure and reduced ref lood cooling.As a result, the no single failure case full ECCS flow produces a higher Peak Clad Temperature (PCT)than the single failure case with the loss of one LPSI pump.Table 1.1 presents the peak cladding temperatures.
and maximum metal-water reaction for the two cases.Figures 2.4 through 2.6, 2.9-2.12, and 2.20 present plotted results of the analyses of the limiting break (1.0 DECLG)applicable for both the single failure and the full ECCS flow cases.Unless otherwise noted on'he figures, time zero corresponds to the time of break initiation.
System blowdown results prior to ECCS initiation apply for both single failure and full ECCS cases.Figures 2.7, 2.8, 2.13-2.19 and 2.21-2.26 provide the remainder of the plotted results for the ECCS single failure case.Figures 2.27-2.41 provide the additional results for the full ECCS flow, no single failure case.
XN-NF-82-35'upplement 1 The peak cladding temperature calculated for the equivalent dou-ble-ended cold-leg guillotine break configuration (Cp=l.0)with single failure is 2091oF at a total linear heat generation rate of 12.0 kw/ft (F;=2.1)for ENC fuel Cycle 4.The maximum local metal-water reaction for this case is 5.91K after 400 seconds, and the total core metal-water reaction reached is less than 1.0X.The peak cladding temperature calculated for the equivalent double-ended cold leg guillotine break configuration (Cp=1.0)with full ECCS flow is 2198oF at a total linear heat generation rate of 11.66 kw/ft (F T=2.04)for Cycle 4.The maximum local metal-water reaction for this case Q=is 7.62K after 400 seconds, and the total core metal-water reaction reached is less than 1.(C, all below the limits set by the criteria of 10 CFR 50.46.The results of the limiting break calculation with single failure (4l are essentially the same as those reported in XN-NF-82-35
)for the break spectrum and exposure analyses.Thus, the net effect of the revised stored energy model and the updated D.C.Cook Unit 2 application on final ECCS results is minimal.
Table 2.1 Donald C.Cook Unit 2 System Input Parameters XN-NF-82-35 Supplement 1 Primary Heat Output, MWt*Primary Coolant Flow, ibm/hr Primary Coolant Volume, ft3 Operating Pressure, psia Inlet Coolant Temperature, oF Reactor Vessel Volume, ft3 Pressurizer Volume, Total, ft3 Pressurizer Volume, Total, ft3 Accumulator Volume, Total, ft3, (each of four)Accumulator Volume, Liquid, ft3 (each of four)Accumulator Pressure, psia Steam Generator Heat Transfer-Area, ft2 Steam Generator Secondary Flow, ibm/hr Steam Generator Secondary Pressure, psia Reactor Coolant Pump Head, ft Reactor Coolant Pump Speed, rpm Moment of Inertia, ibm-ft2 Cold Leg Pipe, I.D.in.Hot Leg Pipe, I.D.in.Pump Suction Pipe, I.D.in'uel Assembly Rod Diameter, in'*Fuel Assembly Rod Pitch, in.Fuel Assembly Pitch, in.Fueled (Core)Height, in'.Fuel Heat Transfer Area, ft2**Fuel Total Flow Area, ft2**3425 142.7 x 106 11,892 2250.542 4,945 1800.1080.1350.950.636 51,500 4(3.685 x 106)820 277.1189.82,000.27.5 29'31.0 0.360 0.496 8.466 144.0 57,327 53.703*Primary Heat Output'sed in RELAP4-EM Model=1.02 x 3425=3493.5 MWt**ENC fuel parameters Table 2.2 Donald C.Cook Unit 2, Cycle 4, DECLG (CD=1.0)Break Events for.Full ECCS Flow, and One LPSI Pump Flow Events Start Initiation of Break Safety Injection Signal Accumulator Injection, Intact Loop Accumulator Injection, Broken Loop End of Bypass (EOBY)Bottom of Core Recovery (BOCREC)Accumulator Empty, Intact Loop Safety Pump Injection Peak Cladding Temperature Reached Time (Seconds)Single Failure of~OS t 0.0 0.05 0.65 15.5 3.3 24.27 40.79 51.27 25.65 282 Full ECCS Flow 0.0 0.05 0'5 15.5 3.3 24.27 40.47 53.02 25.65 341 C/l>C C 2.U 2: ID I B CQ (D fV I c+4l (Jl 5'TEAN GENERATOR PRESSUR ILER STEAN GENERATOR Qs 60 Qlo 39 46 INTACT LOOP Q NOOEL VOLIRIE Q FLOM 3UNCTION REACTOR VESSEL Qi 8ROKEN LOOP 18 18 19 Q46 19 20 20 56 Q9 9 59 Ql'IT LEG PUNP QS 54 Qz is 8REAX LOCATION 15 HOT LEG 13 13 I Il 1 21@42 Q8 li Q39 QS 28 Q32 44 Qi 38 34 CI@49 Q4 053 0>>PUNP 24 43@SO CONTAINMENT
~6 43 (g 48 52 26 26 tos 25 17 21'6 6 Qzi 22 22 55 10 Qz ACCUWLATOR 31 33 Qs 42 Q3 29 lO 36 3 I 31 30 3 Qll ACCIlIILATOR 23 C/l>C C R 0 I I CO fD I e+CrJ Cll Figure 2.1 System Blowdown Nodalization for the Donald C.Cook Unit 2 PWR D.C.COOK 17X17 FUEL FQ~2..04 CYC4 CLAD 00~~360 FZ3*1 31613 5 I ce CS d5.0 47 O.t 0.3 0.4 0.5 O.l RELATIVE HEIGHT X/L Figure 2.2 Axial Peaking Factor versus Rod Length 1Aw 0.0 OoS 14 R X C I'o CD I 2 CI'D PO I M 4J Ql WC.COOK 17X17 FUEL FQ~2.10 CYCI CLAD OD~~580 F2~1~5548i.CD hC<<C 4J Ol X Ooi O.i O.S 0.4 O.S O.C RE'LATXVC HCXt HT X/L Oct 0.0 0+$a.d Vl X C R I'D 2.'D I 3 Co ID PO I C7l Figure 2.3 Axial Peaking Factor versus Rod Length 1.0 OECLG Break with Single Failure ECCS Flow 1~B OLCLC Bl 044OOVN AIIALYSXS FOR O.C.COOXB.4 1VX11 INC FUEL COCCI 14 lC tl tl TIgE AFTER BRCAk (BE'C)Figure 2.4 Upper Plenum Pressure 1.0 DECLG Break{Single Failure and-Full ECCS Flow)=40 C/)0<L Z't3 I'ZI W ID I 3 Co ID M I M 4J Vl I.II DECLC BLOIIOOIII ANALI EB 7ltlt O.L.LDONE.I77!7 ENC BIIEL CICI 1t ZO tl t$TipRFTQR BRCRk f SE'C l Figure 2.5 Pressurizer Pressure 1.0 OECLG Break (Single Failure and Full ECCS Flow)Ch>C C R I CD I R GO ftD I Vl 1 0 QECLQ BLQVQOMN ANALYSXS FOII Q C COOXC, 17X17 ENC FUEL CYCI t lK lC'5 TXHC AFTER HREQK (SE'C)Figure 2.6 Total Break Flow 1.0 DECLG Break Fl w 1 0 DECLG BLOMDOQN~0 C.COOK 2 o AVG.CORE.~.17X17 FUEL CYC4 1R ll t0 tA tl TXME AFTER BREAK (SEC)Figure 2.7 Average Core Inlet Flow L.U Ok(;Lt'reak (Sing le Failure)Cll>C C 2: Cl I'0 R CD I 3 Co CD M I CPl
1.0 DECLt
BLOQDOMH i D.C COOK P.o AVG.CORE i....17X17 FUEL CYC4 12 ll RO K4 t8 TINE AFTER BREAK (SEC)figure 2.8 Average Core Outlet Flow 1.0 DECLG Break (Single Failure)I/I>C C R I'U W ID I B 00 ID I w 4l CJl
1.0 OE'CLC
BLOQOOMjM WALYSXS FOR P.C.COOk2., 17X17 E'NC FUEL CYC$Q 1t Z,O tl tS TlPE AFTER SREQk (SE'C)Figure 2.9 Downcomer Flow Rate, 1.0 DECLG Break (Single Failure and Full ECCS Flow)A K C R I C7 III I CO CD PQ I l0 Vl 1~0 DECLG BLOIIQOWN ANALYSXS fOR OLC.COOKE 1 17X17 EIIC FUEL CYCI 1'K iC KO tl?f TIRE AFTER 8REAK t SEC)Figure 2.10 Pressurizer Surge Line Flow, 1.0 OECLG Break (Single failure and Full ECCS Flow)
VII OSCLG SLOIIOONII AIIALYSSS SOR O.C.COONS.17117 SAC PIIRL CYCI CSS I S 1Z 1C i0 tl tJ'3t Taws'FTm saepi r sec)Figure 2.ll Flow From Intact Loop Accumulators, 1.0 DECLG Break (Single Failure and Full ECCS Flow)IO C/l)C C 2: 0 I CD CO CD I~CpJ Vl
1.0 OECLG
BLOMOOMN ANALYSIS FOR O.C.COOkE 1 17X17 ENC FUEL CYCI S 12 LC t0 t$0t.TXWE At=Ted'R BRE'Ak t SEC)Figure 2.12 Flow From Broken Loop Accumulators, 1.0 OECLG Break (Single Failure and Full ECCS Flow)
1.0 DECLG
SLOWDOWN i O.C.COOK 2 i HOT CHANNEL i 17X17 FUEL CYC4 R N N fA tl TXHE AFTER BRCAK (SEC)Figure 2.13 Hot Channel Average Fuel Temperature
1.0 DECLG
Break (Single Failure)Ch K C R D I D Cb I 00 (D I 40 Ul X.O OCCLt: BLOVOO~V, O.C.COOX a, Hm CHAWaL, uXu FuEL t:VCS 1K 10 tO K4 TXHE AFTER BREAK (SEC)Figure 2.14 Clad Surface Temperature
1.0 DECLG
Break (Single Failure)A OK D I'D CD Co CD co Vl
1.0 OECLG
BL01400MN D.C.COOK P.HOT CHANNEL 11XX7 FUEL CYC4 Ho I-O Pf v Qo 5m tLo o Gc'1S KO t4 tf TXME AFTER BREAK (SEC)Figure 2.15 Oepth of Metal-Water Reaction 1.0 DECLG Break (Single Failure)
1.0 OECLG
BLOWOMH.O.C.COOK R, HOT CHAWEL, X7Xa7 FUEL CYCi 12 N CO H RS TXiE AFTER BREAK (SEC)Figure 2.16 Hot Channel Heat Transfer Coefficient
1.0 OECLG
Break (Single Failure)Ch>C C R CI I D R CD I 3 00.CD I CJl i.o DECLG BLOQDOQN, D.C.COOK R, HOT CHANNEL, i7Xi1 FUEL CYC4 Q, 1S t0 tl Kl TIME AFTER BREAK (SEC)Figure 2.17 Hot Assembly Inlet Flow 1.0 DECLG Break (Single Failure)
Z.o OECLt: BLaVDOW, O.C.COOK Z, HOT CHAWEL, uXu FUEL CYCLE Q.I ZO B 20 TXME'FTER BREAK (SEC)figure 2.18 Hot Assembly Outlet Flow 1.0 DECLG Break (Single Failure) 1 0 OE'CLG CYCL4 G~PEX j ICECON CTP ANALYSIS INTR 3X 3 4R5HVT 4 Rl0 l0 SKO NO 3CO 1RO i00)$0 Z00 2(0 tl0 TINE (SEC)Figure 2.19 ICECOH Containment Back Pressure 1.0 OECLG Break (Single Failure)CA K D I D R CD I B Co CD I Vl O.C-COOK 2 NORMALIZED POWER DXu, i.0 OCCAM C.~"<<4.34aS~W.40 1RO 1CO tN t$0 t$0 TINE (SECONOS)Figure 2.20 Normalized Power 1.0 DECLG Break (Single Failure and Full ECCS Flow)
P.C.COOK 2, REFLOOD 17X17 ENC FUEL 1.0 OECLG C~CL4 34P5 HIT lRO SO 5N t40 tN0 TIHE FROH BOCREC (SEC)Figure 2.21 Core Flooding Rate 1.0 DECLG Break (Single Failure)cnx C 2: I D R CD f 6 Co CD I rl 4)Vl O.C.COOK 2 REFLOOD 17X17 E'HC FUEL 1 0 OECLG CYCL4 3425 HIT QO NO 80 KM LIO TXHE FROH BOCREC (SEC)Figure 2.22 Reflood Downcomer Mixture Level 1.0 OECLG Break (Single Failure)C/l>C C R CI I'0 R ID I B CO ID I N 4)CJl O.C.COOK R REFLOOO 17X17 ENC FUEL 1 0 OECLG Cl'CL4 342.5 HWT ltd lN tN tAO fAO TXHE FRON 80CREC (SEC)Figure 2.23 Ref lood Core Mixture Level 1.0 OECLG Break (Single Failure)400 O.C.COOK 2 REFLOOO 17X17 ENC FUEL 1.0 OECLG CYCL4 3485 HMT<0 uO mO SO t(0 atO TIHE FROH BOCREC (SEC)Figure 2.24 Ref l ood Upper Plenum Pressure 1.0 DECLG Break (Single Failure)C/I X D I ID I ID fV I M CA CJl l' O.C COOK R REFLOOD 17X17 ENC FOEL 1.0 OECLG CYCL4 3425 MQT<0 Xae N0 aOO ZA0 eo TIME FROM SOCREC (SEC)Figure 2.25 Reflood Core Saturation Temperature
1.0 OECLG
Break (Single Failure)9)>C C I CD I Co CD I cjl D.C.COOK UNIT 2 17x17 ENC FUEL 1.0 OECLt, Fq=2 IO 1.PCT HODE (HODE 8)AT S.I2 FTo)RUPTURED HODE (HODE 10 AT 6.15 FT)40o0 80.0 120.0 160.0 Z00.0.Z<0.0 ZS0.0 320.0 TIHE-SECONDS Figure 2.26 TOODEE2 Cladding Temperature vs.Time 1.0 DECLG Break (Single Failure)360-0 CA K C I lD ED I 00 40.0 m cv
1.0 QECLG
8LO'MDOW RHIII YSIS FOR Q.C.COORZ, 1'/X17 EIIC FUEL CYCI 12 1C ZO ZC U~II'FTER SRCPk'.l SE'C)Figure 2.27 Average Core Inlet Flow 1.0 OECLG Break (Full ECCS Flow)1C 40 Ch X C I ID I CO ID I CJl 1.0 DECLC.'LOWOOMH ANALYSIS FOR 0.C.COOX'Z, 17X17 ENC FUEL CYCI Q, 1C t0 TINE AFTER 8REAK (SCC)Figure 2.28 Average Core Outlet Flow 1.0 DECLG Break (Full ECCS Flow)CO l/l X C R'Cl I CI ID I B 00 ID I CJl
1.0 QECLG
BLOMDQMt4, D.C.COOK 8, HOT CHAbtNEL, 17X17 FUEL CVC4 tC xe Z4 tt Txwf AFTER BRE'AK (SEC)Figure 2.29 Hot Channel Average Fuel Temperature
1.0 DECLG
Break (Full ECCS Flow) i.o OECLC BLOQOOQ~.D.C.COOK 2,.HOT CHAHVEL, a~Xzz fuEL CYC4 N.ZO t4 RC TIVE AFTER=BRE'AK f SEC3 Figure 2.30 Clad Surface Temperature
1.0 DECLG
Break (Full ECCS Flow)56
1.4 OECLG
BLOWQOWN, 0 C.COOK 2 HGT CHANNEL 17X17 FUEL CYC4 1E lC t5 TINE AFTER BRCAk f SEC)Figure 2.31 Depth of Metal-Water Reaction 1.0 DECLG Break (Full ECCS Flow)40 t/I OC C R I ID I CO ID I CJl 1-0 DE'CLG BLOMOOVH~O.C-'OOK 2 i HOT CHAHINEL.17X17 FUEL CYCk lC ZO-K4 a0 TXVE AFTER BRCAK t SEC)Figure 2.32 Hot Channel Heat Transfer Coefficient
1.0 OECLG
Break (Full ECCS Flow)l/I X C 13 I CD I R CD CD M I M VJ CJl r
1.0 DE'CLG
BLOMDOll!4, D.C.COOk R, HOT CHAHHEL~17X17 FUEL CYC4 Lt tC t0 t4 tS*TTNE AFTER SRE'AK (SEC 1 Figure 2.33 Hot Assembly Inlet Flow 1.0 OECLG Break (Full ECCS Flow)SC 1~0 DECLG BLOVDOW 1 D C COOK 2 i HOT CHAIINEL 17X17 FUEL CYCI?C Kd KC Zl TIVE AFTER BREAK (SEC)Figure 2.34 Hot Assembly Outlet Flow 1.0 OfCLG Break (Full ECCS Flow)3Z.ld C/l>C C R l3 I R n 1 B 00 ID PV I pt 4)fjl
1.0 DECLG
CYCL4i GAPEX i XCECON CTP ANAL,YSXS XNTR Xm;,3lR5MMT 40 120 1CO 200 210 TIHE'SE'C)240 320 3CO ioo Figure 2.35 ICECON Containment Back Pressure 1.0 OKLG Brrak (Fu)1 FCCS Flow)
D.C.COOk'REFLOOD 17X17 FUEL 1 0 PECLG CYCL4.34Z5 HIT (0 Q.O]CO t00 t40 tSO MO TAHE FROH BOCREC t SEC)Figure 2.36 Core Flooding Rate 1.0 DECLG Break (Full ECCS Flow)400 lh X C R I D Cb.I'B Co CD I r+Ca)CJI I v D.C.COOK 2, RKFLOOD l.7X17 FUEL 1.0 DE'CLC CVCLO 34ZS MWT 40 40 1CO 800 f40 880 TINE FROM BOCREC (SEC)Figure 2.37 Ref lood Downcomer Mixture Level 1.0 DECLG Break (Full ECCS Flow)440 Q.C.COOK 2 REF LOOD 17X17 f UEL 1 0 DE'CLG CYCL4 3$25 HWT l0 lZ.O 1CO 800 Z 40 t.80 TIRE fROH SOCRE.'C (SEC)M,O 300<00 Figure 2.38 Ref lood Core Mixture Level 1.0 DECLG Break (Full ECCS Flow)
D.C.COOK 2.REFt OOD 17X17 FUEL 1.0 DEClC CYCL4 352,5 HIT K H D tu~w 07 Vl tv D Dg K K 4J DQ K 4J D D t4<0 lt0 160 t00 240 K%0 TXNF FROM BOCRE'C f SE'C)Figure 2.39 Ref lood Upper Plenum Pressure 1.0 DECLG Break (Full ECCS Flow)
D.C.COOK 2.REFLOOD 17X17, FUEL 1.0 DECLG CYCL4.3585 HWT$0 lt0 1CO t00 t40 5>0 TAHE FROH BOCREC (SEC)Figure 2.40 Ref lood Core Saturation Temperature
1.0 OECLG
Break (Full ECCS Flow)3to 360 400 C/l M C R'Cl I D R CD I B 00 CD~I rt'J CJl D.C.COOK UNIT 2, 17xl7 ENC FUEL 1~0 DECLC FQ 2..04 4.8 Vl 4J 4J IE 4J)Cl~J 4J 5$Cf.4J CL K 4J D J-cs C9 z.H Cl Qg CJ PCT HOLE (HOOK l'K 4T LCt FTr 1 Ke RtJPTVRE'9 BODE (t409E 11 (IT 1.00 FT l Ct R.o 40 0$0.0 120.0 100.0 K00.0 R40.0 TIVE-SE'CORDS f10.0 MO.O Figure 2.41 TOODEE2 Cladding Temperature vs.Time 1.0 OECLG Break (Full ECCS Flow) 51 XN-NF-82-35 Supplement 1 3-0 CONCLUSIONS For breaks up to and including the double-ended severance of a reactor coolant pipe, the Oonald C.Cook Unit 2 Emergency Core'Cooling System will meet the Acceptance Criteria as presented in 10 CFR 50.46 for Cycle 4 operation with.ENC 17xl7 fuel operating in accordance with the LHGR limits noted in Table 1.1.That is: The calculated peak fuel element clad temperature does not exceed the 2200oF limit.2.The amount of fuel element cladding that reacts chemically with water or steam does not exceed I 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 17K 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.
a 52 XN-NF-82-35 Supplement 1
4.0 REFERENCES
1."Exxon Nuclear Company Evaluation Model EXEM/PWR ECCS Model Updates", XN-NF-82-20(P), Revision 1, August 1982;Supplement 1, March 398~2,!98, 2 1 8 9 Richland, Washington 99352.2."GAPEXX: A Computer Program for Predicti.ng Pellet-to-Cladding Heat Transfer Coefficients", XN-73-25, Exxon Nuclear Company, Inc., Richland, Washington 99357, august 13, 1973.3."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, 1974.4."Donald C.Cook Unit 2 LOCA ECCS Analysis Using EXEM/PWR Large Break Results", XN-NF-82-35, Exxon Nuclear Company, Inc., Richland, 9 99352i23 8 5.U.S.Nuclear Regulatory Commission letter, T.A.Ippolito (NRC)to W.S.Nechodom (ENC), SER For ENC RELAP4-EM Update, March 1979.6."ICECON: A Computer Program Used to Calculate Containment Backpressure for LOCA Analysis (Including Ice Condenser Plants)", XN-CC-39, Revision 1, Exxon Nuclear Company, Inc., Richland,~es>ngton 99352, November 1977.7."Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model Update ENC WREM-IIA, XN-NF-78-30(A), Exxon Nuclear Company, 8."Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model", XN-NF-82-07(P), Exxon Nuclear Company, Inc., Richland, Washington 99392, 82.9.G.N.Lauben,"TOODEE2: A Two-Dimensional Time Dependent Fuel Element Thermal Analysis Program", NRC Report NUREG-75/057, May 1975.
53 XN-NF-82-35 Supplement 1 Issue Date: ll/18/82 DISTRIBUTION M.J.Ades D.J.Braun J.C.Chandler RE E.Collingham G AC.Cooke ST E.Jensen W.V.Kayser J.E.Krajicek J.N.Morgan G.F~Owsley G.A.Sofer R.B.Stout T.Tahvi li D.M.Turner P.J.Valentine G.N.Ward G;F.Owsley/USNRC (35)H.G.Shaw/AEP (5)Document Control (10)