ML20141N186
| ML20141N186 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 12/31/1985 |
| From: | Braun D, Collingham R, Krajicek J SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML17278A619 | List: |
| References | |
| XN-NF-85-139, NUDOCS 8603040460 | |
| Download: ML20141N186 (54) | |
Text
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,y XN-h F 139 I
I W\\l3-2 _OCA - ECCS ANALYS S
{
I VIAP_-GR RESU_TS l
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I XN-NF-85-139
!ssue Date: 12/16/85 8
WNP-2 LOCA-ECCS ANALYSIS MAP'MGR RESULTS I
t Prepared by:
. OC D. M. Braun BWR Safety Analysis
//
I Preoared by:
u, ajicek "BWR Sa ety Analysis
,I G snd
'st-Concur:
RY E. Collingh, Manager BWR Safety An ysis
)
Concur:
/
i J. N. Morgarf, Manager Customer Services Engineering Approve:
,dN' H.X. Williamson, Manager Licensing & Safety Engineering Approve:
.I E
G. L. Ritter, Manager Fuel Engineering & Technical Services naa ERON NUCLEAR COMPANY,INC.
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4 NUCLEAR REGUL Taw 00*avtssiON 0iSCLAIMER R
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i XN-NF-85-139 i
Table of Contents 4
1.0 INTRODUCTION
1 2.0
SUMMARY
................. 3 1
- 3. 0 J ET PUMP 8WR ECCS EVALUATION M0 DEL................................... 6 3.1 LOC A De s c ri p t i o n................................................... 6 3.2 EXEM/BWR Application to WNP-2......................................
7 4.0 ANALYSIS RESULTS....................................................
14 4.1 System Analysis...................................................
14 9
4.2 MAPLHGR Results...................................................
15 5.0 References..........................................................
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List of Tables t
Table Page 2.1 MAPLHGR Results for ENC 8x8 Reload Fuel in WNP-2........................
............... 4 3.1 WNP-2 Reactor System Data..................................
10 -
4.1 LOCA Event Times in Seconds for Limiting Break.............
17 4.2 LOCA BOL Results for Limiting 8reak........................
18 1
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tii xn.NF-85-139 List of Figures Figures Dage 2.1 MAPLHGR vs. Assembly Average Burnup for ENC 8x8 Reload Fuel in WNP-2..................................
5 3.1 System Blowdown Nodalization for WNP-2........................
11 l
3.2 Hot Channel Nodali zation for WNP-2............................
12 3.3 System Refill /Reflood Nodalization for WNP 2..................
13 4.1 Blowdown System Pressure. 0.6 DES /RS Break.........
19 I
4.2 Blowdown Break Flow. 0.6 DES /RS Break........
................ 20 1
4.3 Blowdown Average Core inlet Flow, 0.6 DES /RS Break............ 21 4.4 Blowdown Average Core Outlet Flow.
0.6 DES /P.S Break..............................................
22 4.5 Blowdown HOT CHANflEL Inlet Flow, 0.6 OES/RS Break............................................. 23 I
4.6 Blowdown HOT CHANNEL Outlet Flow, 0.6 DES /RS Break..............................................
24 4
4.7 Blowdown HOT CHANNEL Heat Transfer Coefficient.
0.6 DES /RS Break.................
25 4.8 Blowdown Intact Loop Jet Pump Drive Flow.
f.
0.6 DES /RS Break.......................................
..... 26 4.9 Blowdown intact Loop Jet Pump Suction Flow, 0.6 DES /RS Break..............................................
27 4.10 Blowdown Intact Loop Jet Pump Exit Flow, i
0.6 DES /RS Break.............................................
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iv XN-NF-85-139 4{
1 List of Figures (Cont.)
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Figure Page 4.11 Blowdown Broken Loop Jet Pump Drive Flow, 0.6 DES /RS Break............................................
29 j
1 4.12 Blowdown Broken loop Jet Pump Suction Flow.
O.6 DES /RS Break............................................
30 l
4.13 Blowdown Broken Loop Jet Pump Exit Flow, s
l 0.6 DES /RS Break............................................
31 g
4.14 Blowdown Upper Plenum Mixture Level, 0.6 OES/RS Break............................................
32 4.15 Blowdown Middle Downcomer Mixture Level, 0.6 DES /RS Break............................................
33 3
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4.16 Blowdown Lower Downcomer Mixture Level, l
0.6 DES /RS Break............................................
34 j
j 4.17 Blowdown Upper Plenum liquid Mass, i
0.6 DES /.RS Break............................................
35 I
4.18 Blowdown Upper Downcomer Liquid Mass, l
0.6 DES /RS Break............................................
36 4.19 Blowdown Lower Downcomer Liquid Mass, 0.G DES /R3 0reak......................................
37
'I 4.20 Blowdown Total Lower Plenum Liquid Mass, 0.6 DES /RS Break............................................
38 4.21 Refill /Reflood Lower Plenum Mixture Level, l
0.6 OES/RS Break............................................
39 l
4.22 Refill /Reflood Normalized Liquid Flow at Core Midplane Entrainment, 0.6 DES /RS Break...............................
40 I
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s Xti-NF-85-139 List of Figures (Contj Figure Page I
I 4.23 Blowdown HOT CHANNEL Heat Transfer Coefficient 0.5 DES /RS 8reak............................................
41 4.24 Blowdown HOT CHANNEL Center Volume Quality I
0.6 DES /RS Break............................................
42 4.25 Blowdown HOT CHANfiEL Center Volume Coolant Temperature 0.6 DES /RS Break............................................
43 l
4.26 Typical Hot Assembly Heatup Results, BOL, HAPLHGR =
13.0.........................................
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Xft flF-85-139 lI i
1.0 iflTRODUC T10ft I
The results of a Loss of Coolant Accident (LOCA)
Emergency Core Cooling I
System (ECCS) analysis for Exxon fluclear Company (EflC) 8x8 fuel in the 1
f Supply System fluclear Project flumber 2 (WflP 2) reactor are summarized in this document.
The results of this analysis are presented in terms of the Maximum Average Planar Linear Heat Generation Rate (dAPLHGR) limit as l
l a function of fuel average exposure.
These calculations were performed with the generically approved EflC EXEM/BWR Evaluation Model(1,2)
I according to Appendix K of 10 CFR 50(4), and the results comply with the l
U.
S.
flRC 10 CFR 50.46 criteria.
I j
A BWR $ LOCA break spectre analysis was performed for 100 percent power l
and 106 percen of rated flow using WflP 2 plant data and has been I
i previousif raparted.(3) Th u analysis showed that the limiting break for a B'A 5 using EllC methodology is a split break on the suction side of the l
recirculation pump with an area equal to six tenths of the area of the I
i largest double-ended recirculation pipe break,
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2 Xft-fif 139
,t 2.0
SUMMARY
I A limiting break calculation was performed for the WflP-2 plant specific geometry and conditions to establish boundary conditions for the heatup I
analyses.
This limiting break calculation was performed at a plant power f
level of 3389.46 MWt, which is 102 percent of rated power, and 106 percent of rated core flow to support operation within the WflP-2
(
power-flow region up to an increased core flow of 106?..
The exposure dependent MAPLHGR limit was determined for EflC fuel from beginning of life (80L) to an assembly exposure of 35 GWD/MTM using the Itmiting break boundary conditions. The resulting MAPLHGR limit for EllC j g fuel is presented in Figure 2.1.
The MAPLHGR limit of Figure 2.1 is
'g i
constant at 13.0 kw/ft for exposures from 0 to 20 GWD/MTM and the limit j
decreases approximately linearly from 13.0 to 7.9 kw/ft for exposures j
from 20 to 35 GWD/MTM.
The limit applies to the initial and subsequent reloads incorporating EflC 8x8 fuel of the design analyzed for the WilP 2
}
reactor.
t All of these calculations were performed according to Appendix K of 10
- I t
CFR 50(4) and the MAPLHGR Jf figure 2.1 satisfies the requirements l
specified by 10 CFR 50.46 of the U.S.
Co,de of Federal Regulations.
Operation of the Wf1P 2 reettor with EllC 8x0 fuvi within the limit of
)
Figure 2.1 assures that the Emergency Core Cooling System for the WflP 2 reactor will meet the U.S.
flRC acceptance criteria for breaks up to and including the double ended severance of a reactor coolant pipe.
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3 Xfi-flF 139 I
That is:
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The calculated peak fuel element clad temperature does not exceed 5
the 2200 F limit.
2.
The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1*.' of the total amount of Zircaloy in the reactor.
3.
The cladding temperature transient is terminated at a time the core geometry is still amenable to cooling. The hot fuel rod cladding i
oxidation limit of 17?. is not exceeded during or after quenching.
=
4 The system long term cooling capabilities provided for the previous l
core remains applicable to ErlC fuel, I
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4 Xfi-flF 139 5
Table 2.1 MAPLHGR Results for ENC 8x8 Reload Fuel in WNP-2 I
Assembly Average Local Peak Clad I
(
Burnap MAPLHGR MWR*
Temerature
( )
(F)
(GWD/MTM)
(kw/ft)
~0-13.0
- 0. 8 1765.
5.
13.0 0.48 1766.
I 10.
13.0 0.47 1765.
15.
13.0 0.47 1772.
20.
13.0 0.54 1788.
I 25.
11.3 0.34 1699.
I 30.
9.4 0.17 1521.
I 35.
7.9 0.10 1397.
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- Metal Water Reaction I
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6 y.N-NF-85-139 5
3.0 JET PUMP BWR ECCS EVALUATION MODEL 3.1 LOCA Description I
A Loss-of-Coolant Accident (LOCA) is defined as a hypothetical rupture of the reactor coolant system piping, up to and including the double-ended rupture of the largest pipe in the reactor coolant system or of any line connected to that system up to the first closed valve.
In the unlikely event a LOCA occurs in the WNP-2 plant, reactor system coolant inventory loss would result in a high containment drywell pressure concurrent with low reactor water level. These two events would provide a safety signal which would bring emergency coolant injection systems into operation to limit the accident consequences.
I During the early phase of a LOCA depressurization transient, core cooling is provided by the exiting coolant inventory. During the reactor system depressurization, the High Pressure Core Spray (HPCS) provides additional core heat removal.
In the latter stage of system depressurization and as well as after depressurization has been achieved, the HPCS provides core cooling and the Low Pressure Coolant Injection (LPCI) supplies additional liquid to rapidly refill the bypass region of the reactor vessel and reflood the core.
The liquid which fills the bypass can flow into the lower plenum through the core support structure paths and the bypass holes drilled in the tie plates of the fuel. During the core reflood process, cooling is provided above the mixture level by entrained reflood I
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XN-NF-85-139 I-liquid and below the mixture level by pool boiling processes.
These reflood processes provide heat removal at all core elevations to terminate the core temperature transient rise.
\\
3.2 EXEM/BWR Atolication to WNP-2 The ENC /8WR ccdes were used for the the LOCA ECCS analysis for the WNP-2 reactor. The versions of the EXEM codes used in this MAPLGHR analysis are the same as used in the WNP-2 break spectrum analyses.(3) EXEM/BWR is comprised of the RODEX2, RELAX, FLEX, and HUXY/BULGEX computer codes.
I The initial stored energies for the RELAX blowdown RELAX / HOT CHANNEL and HUXY/BULGEX calculations are determined with the RODEX2 code.
l The RELAX code is used to calculate the reactor system behavior during l
the initial portion of the reactor system depressurization transient.
RELAX predicts mass distribution, core and system therral hydraulics, and break flow rates.
The blowdown calculation provides core boundary l
l conditions for the heatup calculation and reactor coolant system I
1 l
conditions for the initialization of the refill /reflood transient calculation using the FLEX code. The RELAX / HOT CHANNEL heat transfer is used in the heatup model only up to the time when the pressure in the upper plenum decreases to the value which corresponds to rated LPCS flow.
The reactor core is modeled with heat generation rate determined from reactor kinetics equations with reactivity feedback and decay heating
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8 XN-NF-85-139 1
required by Appendix K of part 10 CFR 50 to the U.S.
Code of Federal Regulations. For the blowdown calculation, the reactor coolant system is nodalized into control volumes representing reasonably homogenous regions interconnected by junctions as shown in Figure 3.1.
This nodalization is consistent with that used in the BWR 5 break spectrum analysis.(3) Pump performance curves characteristic of the recirculation pumps are used in the analysis.
For the maximum power fuel assembly, a separate RELAX / HOT CHANNEL calculation is used to calculate the cladding-to-coolant heat transfer coefficients and the coolant thermodynamic properties. The HOT CHANNEL analysis uses time dependent plenum boundary conditions from the RELAX blowdown calculation.
The calculated results from the HOT CHANNEL calculation are used as input data to the subsequent hot assembly heatup calculation.
Conservative heat transfer coefficients and fluid thermodynamic properties for the heatup calculation are assured by using the maximum stored energy in the HOT CHANNEL calculation for the generation of this information.
The HUXY/BULGEX hot assembly heatup calculation computes the fuel temperature transient from its initiation through PCT. The RELAX / HOT CHANNEL nodalization is shown in Figure 3.2 and is consistent with that used in the BWR 5 break spectrum analysis.(3)
The FLEX system refill /reflood analysis predicts the latter segment of the reactor coolant system depressurization, the refilling of the lower plenum and the reflooding of the core. The time of hot-node reflood is I
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9 XN-NF-85-139 5
determined by FLEX and is an input quantity to the hot assembly HUXY/BULGEX heatup calculation.
The FLEX system refill /reflood nodalization is consistent with that used in the BWR break spectrum analysis (3) and is shown in Figure 3.3.
The HUXY/BULGEX heatup calculation computes the entire LOCA temperature transient and uses:
fuel stored energy, thermal gap conductivity and dimensions from RODEX2 as a function of power and exposure; time of rated spray, decay power, heat transfer coefficients and coolant thermodynamic properties from RELAX; and time of hot-node reflood and time of bypass reflood from FLEX. These data are input to HUXY/BULGEX to determine the peak clad temperature (PCT) and the cladding oxidation percentage. Bounding fission and actinide product decay heat obtained with end-of-cycle neutronics in the system blowdown calculation assure that the power input to the HUXY/BULGEX hcatup calculation is conservative. Conservative heat transfer coefficients and fluid thermodynamic properties for the heatup calculation are assured by using the maximum stored energy over the life of the ENC 8x8 fuel in the HOT CHANNEL calculation for the generation of this information. The stored energy used in this analysis is more conservative than used in the break spectum calculations (3) and thus higher PCT's are anticipated in this MAPLHGR analysis.
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'I Table 3.1 UtlP-2 Reactor System Data Primary Heat Output, MWt 3389.46 (102*; of rated)
Total Reactor Flow Rate, Mlb/hr 115.01
'I (106% of rated) l
{
Active Core Flow Rate, Mlb/hr 102.36 i
tiominal Reactor System Pressure, psia 1031.
Reactor Inlet Enthalpy, Btu /lb 528.4 Recirculation Loop Flow Rate, Mlb/hr 16.65 Steam Flow Rate. Mlb/hr 14.59 Feedwater Flow Rate, Mlb/hr 14.55 Rated Recirculation Pump Head, ft 8&
Rated Recirculation Pump Speud, rpm 1782 Moment of Inertia, lbm-ft**2/ rad 22,700 Recirculation Suction Pipe I.D., in 21.56
}
l Recirculation Discharge Pipe I.D., in 21.56 18 5
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32 ll l
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Steam Dome f6 Feec*ater g
Upper Downcomer 35 34 HPC3 7, j
--J f $s Ph ;s LPCS l
E ag Upper Pienum 8
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,e JUNCTION Pump Recirculation Recirculation Pump Flow Control Flow Control Valve Valve Fioure 3.1 System Blowdown Nodalization for WNP-2 XN-CH-0508
!l 12 XN-NF-85-139
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UPPER PLENUM g
,g I
"6 "s b l
E a
m4 g 4
E5 "a
g 2
g "i
g LOWER PLENUM g
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JUNCTION
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I 13 XN-NF-85-139 l
l Steam Dome & Upper Downcomer I
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(>2 9
c.
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m Recirctdation e
,k Discharge Piping l
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b h
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n n
g ti
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Lower Plenum oa s
$1
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=
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Figure 3.3 System Refill /Reflood Nodalization for WNP-2 3;
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14 XN-NF-85-139 I
4.0 ANALYSIS RESULTS B
4.1 System Analysis An ECCS analysis calculation of the limiting break has been performed for the WNP-2 plant, and the calculated results are summarized herein.
The break spectrum calculations for WNP-2 have been previously performed and reported. The limiting break was found to be a split break on the suction side of the recirculation pump with a break area equal to sixty percent of the double ended break flow area.
I The limiting calculation for WNP-2 has been performed for 100% power and 106% of rated core flow to contribute to the support of operation within the WNP-2 power-flow region up to an increased core flow of 106 of rated for the WNP-2 reactor. The 106% of rated core flow represents the maximum core flow which the plant can operate while maintaining 100*;
power. The performance of the ECCS analysis at the maximum core flow results in a highest radial peaking factor given that the calculations i
I are initialized the same MAPLHGR and MCPR limits.
Therefore, this analysis covers at lower flow conditions at rated core power because the calculations would be initialized with a lower radial peaking factor while maintaining the same initial MAPLHGR and MCPR values.
I The LOCA system behavior is determined by the system geometry and break size. Core parameters have only a secondary effect on system event I
Il 15 XN-NF-85-139 I
times. For these reasons, the system analysis described in this report is applicable to future cycles of WNP-2 unless system modifications or reyiud operating conditions negate the plant conditions used herein.
The calculated LOCA event times and results for a limiting break occuring when the reactor is operating at these conditions and the MAPLHGR results are summarized in Tables 4.1 and 2.1.
I Major event times and results for the limiting break for WNP-2 are shown as follows: Figures 4.1 through 4.20 are system blowdown parameters, Figures 4.21 through 4.22 are system refill /reflood parameters.
4.2 MAPLHGR Results I
The MAPLHGR analysis PCT results for the WNP-2 reactor are based on the limiting break calculation reported in Section 4.1 and repeated in Reference 3.
This system analysis used plant specific data applicable to WNP-2.
MAPLHGR results are obtained using these LOCA systerr analysis boundary conditions but require an additional RELAX / HOT CHANNEL calculation and a series of HUXY/BULGEX calculations at various fuel exposeres.
l A bounding HOT CHANNEL calculation has been performed for this MAPLHGR analysis in which the fuel stored energy was the maximum for the exposure range of interest for ENC 8x8 fuel.
This bounding HOT CHANNEL calculation provides heat transfer coefficient, fluid temperature, and-I
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16 Xfi-N F 139 I
fluid quality at the plane of interest for the HUXY/BULGEX calculations.
The HOT CHANilEL heat transfer coefficient for the break spectrum calculation in reference 3 is shown in Figure 4.7.
These HOT CHANNNEL calculated parameters for the bounding case are shown in Figures 4.23 through 4.25.
Figure 4.26 shows a typical clad temperature trace as calculated by the HUXY/BULGEX code. Since the same bounding hot channel conditions are used for all of the heatup calculations in the MAPLHGR analysis including the BOL case, the PCT for the BOL case in the MAPLHGR analysis is 67 degrees higher than the BOL case reported in the BWR 5 break spectrum analysis.(3)
I The HUXY/BULGEX calculated results and the corresponding MAPLHGR limits for ENC 8x8 reload fuel are shown in Table 2.1 and Figure 2.1.
These results conform to the U.S.
NRC requirements specified by 10 CFR 50.46.
For the average burnup of the hot assembly, Table 2.1 shows the MAPLHGR, local metal-water reaction and peak cladding temperature.
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Table 4.1 LOCA Event Times in Seconds for Limiting Break I
Start C.00 Initiate Break 0.05 Feedwater Flow Stops 0.55 Steam Flow Stops 5.05 Low Mixture Level 10.1 Jet-Pumps Uncover 11.8 Recirc. Suction Nozzle Uncovers 16.7 1
Lower Plenum Flashes (Quality >0) 18.2 l
HPCS Flow Starts 18.5 LPCI Flow Injection Starts 62.3 g l ll Rated Spray Calculated 74.4 Depressurization Ends (vessel E
pressure reaches 1 atm.)
126.0 5
Start of Reflood (high density a
fluid enters core) 162.2 E
Peak Clad Temperature Reached 169.0 I
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Table 4.2 LOCA BOL Results for Limiting Break Reactor Operating Conditions 100% Rated Power /
Result 106% Rated Flow Peak Cladding Temperature, OF 1765 Peak Temperature Axial Location, ft 6.25 Local Zr/ Water Reaction (Max), %
0.49 4
Total Hydrogen Generatied, % of Total Zr Generated
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45 XN-NF-85-139
5.0 REFERENCES
1.
" Generic Jet Pump BWR 3 LOCA Analysis Using the ENC EXEM Evaluation I
Model," XN-NF-81-71( A), Supplement 1, Exxon Nuclede Company, Sep-tember 1982.
I 2.
" Generic LOCA Break Spectrum Analysis for BWR 3 and 4 with Modified Low Pressure Coolant Injection Logic," XN-NF-84-ll7(P), Exxon Nuclear Company, December 1984.
3.
"LOCA Break Spectrum Analysis For A BWR 5," XN-NF-85-138(P), Exxon Nuclear Company, December 1985.
I
" Acceptance Criteria for Emergency Core Cooling Systems for Light 4.
Water Cooled Nuclear Power Reactors," 10 CFR 50.46 and Appendix K of 10 CFR 50, Federal Register, Volome 39, Number 13, January 4, 1974 I
I I
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XN-NF-85-139
}I Issue Date: 12/16/85 l
i i
WNP-2 LOCA-ECCS ANALYSIS MAPLHGR RESULTS j
Distribution 0.J. Braun lI J.C. Chandler R.E. Collingham J.B. Edgar J.G. Ingham 4
f S.E. Jensen J.E. Krajicek l
J.N. Morgan E
0.R. Swope 3
H.E. Williamson i
WPPSS/J.B. Edgar (22)
Document Control (5) i I ilI 4
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