ML19347B619
| ML19347B619 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 04/24/1980 |
| From: | Collingham R, Morgan J, Sofer G SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML19347B615 | List: |
| References | |
| XN-NF-79-095, XN-NF-79-095(NP)-R01, XN-NF-79-95(NP)-R1, NUDOCS 8010150452 | |
| Download: ML19347B619 (55) | |
Text
_ _ _ _ _ _ _
XN-NF-79-95(NP)
W Revision 1 04/24/80 ECCS EVALUATION OF DRESDEN-I USING THT cXXON NUCLEAR COMPANY WREM-BASED NON-JET PUMP EVALUATION MODEL Drafted by: R. D. Hyman D. J. Braun G. C. Cooke f
J. E. Krajicek i
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Prepared by:
() ) (,.Ys'ac hen 1/? >/.10 s
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_z R. E. Collingham; Manager Systems Model Development i
Approved:
1jku c.u pc
,3.2[, ' b7' J),N. Morgan,MWnager Licensing & Safety Engineering J
Approved:
M YV8-fc; uc ea Eg ering l
l ERON NUCLEAR COMPANY,Inc.
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NUCLEAR REGULATORY COMMISSION DISCLAIMER l
I lMPORTANT NOTICE REG ARDING CONTENTS AND USE OF THIS DOCUMENT _
PLE ASE READ CAREFULLY This technical report was derived through research and development programs sponsored by Ex xon Nuclear Company, Inc.
It is being sub i
mitted by Exxon Nuctear to the USNRC as part of a technical contri-bution to facilitate safety analyses by licensees of the USNRC which utilize Exxon Nuclear fabricated reload fuel or other technical services provided by Exxon Nuclear for licht water power reactors and it is true and correct to the best of Exxon Nuclear's knowledge, informa tion, and belief. Tt information contained herein may be used by the USNRC in its review of this report, and by licensees or applicants before the lj USNRC which are customers of Exxon Nuclear in their demonstration of comoliance with the USNRC's regulations.
E Without derogating from the foregoing, neither Exxon Nuclear nor any person acting nn its behalf.
l I
A.
Makes any warranty, express or implied, with respect to i
the accur acy, completeness, or usefulness of the infor-mation contained in this document, or th at the use of 3 ]
an y information, apparatus, method, or process disclosed in this document will not infnnge privately owned rights; i
or
'l 8,
Assumes any liabilities with respect to the use of, or for darrages resulting from the use of, any information, ap-paratus, method, or process disclosed in this document.
i 1ll XN-NF FOO, 766 Il I
-i-XN-NF-79-95(NP)
Revision 1 f
f TABLE OF CONTENTS Section Page
1.0 INTRODUCTION
AND
SUMMARY
1 2.0 ECCS EVALUATION MODELS FOR DRESDEN-1.........
7 2.1 SYSTEM BLOWDOWN MODEL..............
7 2.2 EMERGENCY CONDENSER MODELING 8
2.3 HEATUP ANALYSIS MODELS 9
3.0 DRESDEN-1 BREAK SPECTRUM RESULTS...........
15 l
3.1 BREAK LOCATION RESULTS 15 3.2 BREAK SIZE AND CONFIGURATION RESULTS 16 3.3 NON-RECIRCULATION BREAKS 18 40 HEATUP ANALYSIS RESULTS 36 5.0 MODEL HISTORY 41 5.1 RELAP4-EM GENE 0 LOGY...............
41 q
l 5.2 MODEL CHANGES FOR THE DRESDEN-1 ANALYSIS 42
{
6.0 CONCLUSION
S 46
7.0 REFERENCES
48
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-ii-XN-NF-79-95 (NP)
Revision 1 LIST OF TABLES Table Page 1.1
SUMMARY
OF RESULTS FOR LIMITING BREAK.............
3 2.1 INITIAL OPERATING DATA USED IN ANALYSIS...........
12 3.1 BREAK LOCATION STUDY
...................20 3.2 LARGE BREAK EVENT TIME RESULTS
...............21 3.3 LARGE BREAK HEATUP RESULTS
.................22 3.4 SMALL BREAK EVENT TIME RESULTS
...............23 3.5 SMALL BREAK HEATUP RESULTS
.................24 3.6 NON-RECIRCULATION BREAK RESULTS...............
24
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4.1 HEATUP ANALYSIS RESULTS
SUMMARY
38 4.2 AXIAL POWER PEAK LOCATION SENSITIVITY STUDY
........39 l
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-lii-XN-NF-79-95( NP)
Revision 1 r
LIST OF FIGURES Figure Page 4
L 1.1 DRESDEN-1 MAPLHGR CURVE 5
1.2 DRESDEH-1 LARGE BREAK SPECTRUM l
1.3 DRESDEN-1 SMALL BREAK SPECTRUM 6
8 2.1 BLOWDOWN SYSTEM N0DALIZATION FOR DRESDEN-1 13 L
2.2 H0T CHANNEL N0DALIZATION.
14 L
r 3.1 PRESSURE IN UPPER PLENUM...
25 L
3.2 BREAK MASS FLOW 26 I
3.3 HOT CHANNEL INLET MASS FLOW 27 s
3.4 H0T CHANNEL EXIT MASS FLOW..............
28 I
s 3.5 AVERAGE CORE INLET MASS FLOW........
29 r
3.6 AVERAGE CORE EXIT MASS FLOW 30 L
3.7 HEAT TRANSFER COEFFICIENT IN HOT N0DE 31
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3.8 STEAM DRUM MIXTURE LEVEL.
32 3.9 QUALITY IN HOT N0DE 33 e
L 3.10 HPCI MASS FLOW.
34 3.11 TOTAL MASS IN UPPER PLENUM.......
35 r
L 4.1 HEAT TRANSIENT-ENC FUEL-BOL 40
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XN-NF-79-95(NP)
Revision 1
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1.0 INTRODUCTION
AND
SUMMARY
This document presents results of LOCA analyses performed for the Dresden-1 reactor with ENC WREM-Based NJP-BWR ECCS Evaluation Model.II)( )The results support an ECCS allowable Maximum Average Planar Linear Heat Generation Rate L
(MAPLHGR) of 13.4 kw/ft for ENC 6x6 fuel assemblies at beginning-of-life.
The
(
exposure dependence of the MAPLHGR limit for both ENC and the already irradiated i
Gulf United Nuclear fuel is shown in Figure 1.1.
These MAPLHGR limits corres-f' pond to the liniting 3.40 ft split break between the recirculation pump and 2
the steam generator.
An example problem analysis of a hypothetical large break was previously performed using the ENC evaluation model and reported to the NRC Staff.(3)
The break spectrum results presented herein identify the location, size, and
(
configuration of the worst break in the Dresden-1 reactor system in conform-ance to the requirements of 10 CFR 50 Appendix K and the criteria specified in 10 CFR 50.46.I4) The analyses explicitly consider the new HPCI Dresden-1 f
system as designed for injection into the upper plenum.
Loss of one of the two HPCI pumps is assumed to be the worst active-single-failure in all analyses.
(
Three (3) recirculation line break locations were analyzed in order to identify the worst break location.
The locations evaluated were:
(1) On the suction side of the pump.
(2) Between the pump and the steam generator (3) Between the reactor vessel and the steam generator.
(
4 l XN-NF-79-95(NP)
Revision 1 L
l 1
l The break location analyses showed the most limiting break location
[
l to be between the pump and the steam generator (item 2 described above).
Non-recirculation line breaks were also analyzed.
These non-recircula-tion line breaks were shown not to be limiting.
A spectrum of break sizes (0.1 ft to 4.2 f t ) and configurations were j
analyzed at the limiting break location and the results of this analysis are i
summarized in Figures 1.2 and 1.3.
The worst or limiting break was determined j
to be 3 an ft split break.
Table 1.1 gives the calcelated results for this limiting 3.40 f t split break.
These results are t,ased on a center-peaked axial power profile for i
ENC Type XN-1 fuel at beginning-of-life (BOL) fuel conditions, and a MAPLHGR i
j of 13.40 kw/ft.
t j
The following sections of this report present details of the identifi-cation of the limiting break and point out special model features required l
for this analysis which are related to the Dresden-1 specific plant design.
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9, XN-NF-79-95 (NP)
Revision 1 TABLE 1.1
SUMMARY
OF RESULTS FOR LIMITING BREAK 2
3.40 ft split configuration Beginnir.g-of-Life ENC fuel MAPLHGR = 13.4 kw/ft PCT 2024 F Time of PCT 335 sec
% Local Metal Water Reaction (MWR) 4.6%
% Core Wide MWR
<1.0%
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14 EIlC and Gulf United fluclear fuel with 35 active rods h
13 N
Gul f United fluciear fuel 3
12 with 36 active rods
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11 10 k
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10,000 20,000 30,000
{y O N Burnup, W D/IIII!
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Figure 1.1 Dresden-1 f1APLHGP Curve
"; _E*1_" _ _ " _ " - "_ _ i
2 CONTRACTION COEFFICIENT FOR 4.25 Ft GUILLOTINE BREAKS 0.1 0.2 0.a 0.6 0.8 1.0 I
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O PEAK CLADOING
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N BREAK SIZE
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b = GUILLOTINE BREAK LARGE BREAKS LIMITING LOCATION
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_e BREAK AREA FOR SPLIT BREAKS IN Ft 1
5 Figure 1.2 Dresden-1 Large Break Spectrum
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. XN-NF-79-95(NP)
Revision 1 2.0.ECCS EVALUATION MODELS FOR DRESDEN-1 2.1 SYSTEM BLOWDOWN MODEL The ENC WREM-based NJP-BWR ECCS Evaluation Model uses the RELAP4-EM code to predict the space and time variations of the thermal-hydraulic conditions of the reactor during the LOCA. RELAP4-EM version ENC 28F* was used for large break analyses.
For the large break analyses, the Dresden-1 plant is modeled as in the example problem ( ) using 57 volumes, 70 junctions, and 50 heat slabs.
l The blowdown nodalization for large breaks is shown in Figure 2.1.
All volumes in the large break model are homogeneous with the exception of the steam drum volume.
No credit was taken for the emergency condenser in the recirculation line large and small break analysis, i.e., valve 66 in Figure 2.1 not opened.
For the recirculation line small break analyses, the axial core nodes for the average core (Volumes 3,4,5) and the hot channel (Volumes 6 through 12) were combined and represented by une fluid volume as required by the ENC NJP-BWR small break model.(I)
RELAP4-EM version ENC 28J* was used for the recirculation line small break analyses. This version has an improved small break heat con-ductor representation as described in Section 5; this small break model irgrovement would have no impact on large break results.
The LPCS line break was also run with this small break model.
The transition break size for large/small break transition was selected on the ratio of break size to primary system volume as in the previous applica-tion of the ENC NJP-BWR small break model to Big Rock Point.(0) This method 2
I calculates a transition break area of 0.75 ft for Dresden-l.
l
- For description of RELAP4-EM versions, see Section 5.0.
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I i XN-NF-79-95 ( NP)
Revision 1 The HPCI break was analyzed with RELAP4-EM version ENC 28N" using the
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above described small break model, except with the emergency condenser included.
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This RELAP4 version contains natural circulation and condensation heat transfer i
correlations to model the emergency condenser.
The blowdown analysis begins from the assumed initial operating condition of 102% of rated power and continues through pipe rupture and system decom-pression through the onset of emergency coolant flow until rated core spray is reached.
Initial operating conditions are listed in Table 2.1.
The worst single failure was taken to be loss of one HPCI pump.
l The break spectrum analysis (results shown in Figures 1.2 and 1.3) was performed with the volume above the turning vane assumed to be two phase fluid and with a 10 second emergency diesel generator startup time.
Review I
1 of plant operating data showed the volume above the turning vane to be filled 1
l with saturated steam rather than two phase fluid.
The limiting break was rerun with saturated steam above the turning vane and a more conservative diesel startup time of 12 seconds. With these changes implemented, the PCT 1
i was shown to decrease by 27 F.
This updated limiting break blowdown run was used for the heatup analyses described in Section 4.0.
2.2 EMERGENCY CONDENSER MODELING The emergency condenser was assumed to operate during the HPCI line break.
The RELAP4-EM model of the emergency condenser (including ENC 28N con-i densation correlations) was confirmed to yield the design heat removal of the I
emergency condenser.
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XN-NF-79-95 (NP) l
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l Revision 1 A sensitivity study was performed on the effect of the emergency condenser during large breaks.
The ir. fluence of the emergency condenser l
was shown to have a small effect during large breaks (less than a 15 F l
PCT increase).
Conservatisms in the detailed heatup analyses of Section l
4.0 along with the calculated margins to 10 CFR 50.46 limits in the heatup calculations are sufficient to absorb the impact if the emergency condenser were to be included in the limiting large break BLOWDOWN calculation.
2.3 HEATUP ANALYSIS MODELS The ENC WREM-Based NJP-BWR ECCS Evaluation Model invo'ves RELAP4-EM/
HOT CHANNEL and HUXY( )/BULGEX(10) calculations for the heatup analysis.
The HOT CHANNEL calculations establish convective heat transfer conditions during blowdown for the hot assembly.
These condi.tions are then used in HUXY/BULGEX calculations for the detailed thermal response of the fuel assembly in the postulated LOCA.
The HUXY/BULGEX calculations provide the peak clad tempera-ture (PCT) and the maximum local metal water reaction (MWR) that occur in the LOCA.
The~ RELAP4-EM/H0T CHANNEL calculations were made both as an integral part of the RELAP4-EM/ BLOWDOWN calculation and as a separate calculation using time dependent boundary conditions above and below the hot assembly from a RELAP4-EM/ BLOWDOWN calculation.
The HOT CHANNEL model represents a single limiting (hot) fuel assembly over its full active length.
Nodalization of the hot assembly in both RELAP4-EM/ BLOWDOWN calculations or separate HOT CHANNEL calculations is the same and is shown in Figure 2.2.
This nodalization divides the active fuel j
region within the assembly canister into 7 axially stacked fluid volumes l
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i XN-NF-79-95 (NP)
Revision 1 1
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and 7 corresponding heat slabs representing the active fuel rods in the 6x6 a ssembly.
This nodalization is the same as in the sample problem (I)
Heatup analyses to establish the break spectrum for Dresden-1 were based on hot channel fluid conditions in the RELAP4-EM/ BLOWDOWN calculations.
Separate RELAP4-EM/ HOT CHANNEL calculations were made to establish the MAPLHGR limits for ENC and Gulf United Nuclear fuels.
Blowdown boundary n,g!
i 2
j conditions for the ENC limiting 3.40 ft split break were used in these 4
calculations.
These calculations were made with the axial-times-radial peaking that corresponds to the MAPLHGR limits in the present ECCS analysis.
The HOT CHANNEL calculations accounted for fuel type differences and for the I
j effects of fuel exposure.
Fuel stored energy and the variation of stored Ei I
3 j
energy with exposure were calculated in accordance with ENC's approved BWR Fuel Densification Model with the GAPEXUI) computer code.
The NRC enhanced j
fission gas release model for peak pellet burnups in excess of 20,000 MWD /MTM was incorporated into the GAPEX calculations.
4
)
Convective heat transfer coefficients, coolant temperatures and I
fluid quality versus time for the limiting axial node in the HOT CHANNEL i
calculations were input directly into subse'quent HUXY/BULGEX calculations l
for the detailed thermal response of the fuel.
The RELAP4-EM/ HOT CHANNEL calculations were made with RELAP code versions ENC 28F and ENC 28K.
The latter code version has a modified pellet-to-clad gap iteration scheme to I
l improve initial stored energy convergence.
The code updates for ENC 28K are l
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( XN-N F-79-95( NP)
Revision 1 related to the numerical convergence algorithm and not to the gap conductance model itself. This update is not used during the transient calculations.
The HUXY/BULGEX fuel thermal response calculations covered the time period from break initiation until the core spray quenches all the rods at the axial plane of interest in hot assembly.
(
In the HUXY heatup calculations, fuel rod convective transfer between break initiation and the time of rated spray is from the HOT CHANNEL calculation. After the time of rated spray,10 CFR 50 Appendix K spray cooling heat transfer coefficients were applied. The APR79 version of HUXY/
{,,,,.
BULGEX computer code was used in the present Dresden-1 application.
This
(
code version is the same as used in prior ENC WREM-Based NJP-BWR Evaluation Model applications with the exception that a minor update has been made in f
1 Subroutine MWR. Specifically, the update changes the zircaloy weight density in this subroutine from.231 lb/in to.237 lb/in in order to be fully con-l 3
3 sistent with the properties cited in XN-CC-33A(5)
The effect of this change
(
in a Dresden-1 sample calculation was less than a 5 F increase in PCT.
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_. XN-NF-79-95 (NP) l Revision 1 I
Table 2.1 Initial Operating Data Used in
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Analysis System Data Parameter Value Reactor Power, MW 714*
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Reactor Flow, lb/sec 7,083.
Flow Through Average Core, lb/sec 6,502.
Flow Through Hot Channel, lb/sec 14.2 I
Flow Through Core Bypass, lb/sec 566.7 i
j Reactor Inlet Enthalpy, Btu /lb 4 81. 9 Reactor Outlet Enthalpy, Btu /lb 577.5 l
Reactor Outlet Pressure, psia 1,015.0 Steamdrum Pressure, psia 990.0 l
Steamdrum Temperature, F
543.4 i
Primary Steamflow, Ib/sec 444.4 Primary Feedwater Flow, lb/sec 444.4 Primary Feedwater Temperature, F
416.8 Secondary Steamflow, lb/sec 396.7 j
Secondary Feedwater Floyt, lb/sec 396.7 i
j Secondary Feedwater Temperature, F
417.35 Steam Generator Shellside Pressure, psia 510.0 l
Steam Generator Shellside Temperature, F
469.
Fraction of Heat Generated in Fuel 97f HPCI water temperature ("F) 120.0 LPCS water temperature ( F) 100.0 2.51**
p xp Z
r
- 102% of rated power
- Axial (F ) times radial (F ) nuclear peaking factor I
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! XN-NF-79-95 (NP)
Revision 1 3.0 DRESDEN-1 BREAK SPECTRUM RESULTS The purpose of the break spectrum analyses was to establish the l
worst or limiting break case.
The analyses consist of break location, break size and break configuration calculations. The detailed heatup analyses covering different fuel types and fuel exposures was performed for the limiting break case.
These heatup analyses (described in Section 4.0) establish plant operating limits (MAPLHGR's) for the worst break condition which assures conformance with the LOCA-ECCS criteria of 10 l
CFR 50.46 (4) l l
l 3.1 BREAK LOCATION RESULTS The break location sensitivity study includes three possible break locations in the Dresden-I recirculation line.
The breaks were l
evaluated as double-ended guillotine breaks with discharge coefficient l
l of 1.0 for the following locations:
1 (1) The hot leg on the suction side of the recirculation pump.
(2) The piping between the recirculation pump and the steam generator.
(3) The cold leg between the steam generator and the reactor vessel.
Cladding temperature results are given in Table 3.1 and the break between the pump and the steam generator is shown to be limiting.
This is the limiting location because the pump and the steam generator, both major resistance components, are on opposite sides of the break causing the core flow to approach stagnation.
The other break locations have both j
major resistance components in the same flow path creating either a net
l i
. XN-NF-79-95(FP) i Revision 1 5
5'
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j positive or negative core flow.
The results of the heatup calculations for g
l the break location sensitivity study are reported as peak clad temperature g!
i (PCT) differences from the limiting break location value.
l 3.2 BREAK SIZE AND CONFIGURATION RESULTS Various large and small break sizes with both guillotine and split l
configurations were analyzed for the identified limiting location.
Doubl e-ended I
l guillotine breaks were calculated with discharge coefficients of 1.0, 0.8, 0.6, I
and 0.4.
Large split breaks were analyzed with break areas of 4.25, 3.40, 2.53, i
2 l
1.69, and.75 ft.
Small break calculations were performed for 0.75, 0.50, 2
0.25 and 0.10 f t break areas.
The peak cladding temperature (PCT) and metal 2
water reaction results are presented in Tables 3.3 and 3.5.
The 3.40 ft spli t break is shown to be the most severe or limiting break.
Figures 3.1 through 3.11 2
i are for the 3.40 ft split break of the break spectrum analysis (not the rerun as discussed in Section 2.1).
The calculated event times for the break spectrum are also presented in Tables 3.2 and 3.4.
Coincident with the break initiation, offsite power is I
assumed lost.
Loss of offsite power causes the recirculation and feedwater pumps to coast down.
Loss of load causes the primary and secondary steam flows
)
to stop by closure of steam stop valves in half a second.
The ECCS initiation signal comes from either a high containment pressure of 16.7 psia (+2 psig) or 4
I a low liquid level in the steam drum of -24 inches below normal water level.
i I
This ECC signal starts emergency diesel generators and the high and low i
pressure ECC pumps.
The High Pressure Coolant Injection (HPCI) pump is up to speed 17 seconds (19 seconds in the limiting break rerun as discussed in Section 2.1) after the ECCS initiation signal, the Low Pressure Core Spray (LPCS) pumps are up to speed in a maximum of 37 seconds after ECCS I
I XN-NF-79-95 (NP)
Revision 1 trip. The HPCI injection valves require a icw feactor water level of 43 inches above the core and the HPCI pumps at.naximum rated pressure before they can begin opening.
The HPCI injection valves cpen within a 15 second opening time.
Rated HPCI flow is reached when the valves are 25% open (4 seconds after HPCI injection valves begin opening). The LPCS injection valves require a low reactor water level of 43 inches and a pressure drop (AP) of 30 psid across the LPCS valve before it can begin to cren. The 30 psid AP trip corresponds to a reactor vessel upper plenum pressure of 234 psia with the LPCS pumps at rated conditions.
The LPCS injection valve opens with a 30 second opening time.
Rated LPCS flow is reached in 7.5 seconds (valve 25%
open) after the valves begin opening and reactor upper plenum pressure is 155 psia.
Tables 3.2 and 3.3 show that PCT and event times do not vary much with different large break sizes and configurations.
This is due to the major flow resistance of the pump and steam generator.
These major resistances limit the flow to the break and, thus, control system blow-2 down rate.
For breaks smaller than 1.69 ft, the break size limits break flow.
System behavior is then dominated by the break.
Small break PCT and metal water reaction results are presented in Table 3.5.
PCT decreases rapidly with decreasing break size for breaks 2
smaller than 0.75 ft because the HPCI flow soon becomes large in com-parison with the break flow.
The large HPCI flow prevents the core from uncovering for an extended period of time for small breaks.
XN-NF-79-95(NP)
Revision 1 I
3.3 NONRECIRCULATION LINE BREAKS Breaks that are not in the recirculation piping which could result in loss of primary coolant were considered.
These include primary steam line, primary feedwater line, high pressure coolant injection (HPCI) line, and low pressure core spray (LPCS) line. A break in the primary steam line or primary feedwater line is considered to have been enveloped by the recirculation line break location sensitivity study.
As described in Section 3.1, the break location study shows pump suction breaks to be non-limiting due to a net positive core flow.
Steam line and feedwater line breaks are effectively in the recirculation system as they connect directly to the steam drum which is in the pump suction line.
A break in these lines would cause positive core flow as in the pump suction line break.
Thus, steam line and feedwater line breaks would not be limiting.
I A break in the LPCS line c the HPCI line is not covered by the break location study because a break in these systems affects avail-ability of ECC systems.
These breaks were analyzed in detail and shown not to be limitin1 The ENC small break model was used in these analyses.
In the analysis of the LPCS break, no credit was taken for the E
emergency condenser.
The core started to uncover; at 315 secones, however, the HPCI pumps soon filled the upper plenum with water whose head was sufficient to force this water into the core in a homogeneous flow manner (no slip flow) and refill the core. A separate slip-counter-current flow limitation (CCFL) calculation was performed which confirmed I
I I'
XN-N F-79-95( NP)
Revision 1 that water would flow into the core and refill it even sooner if slip /CCFL were considered.
The analysis showed HPCI capable of keeping the core covercd after 375 seconds.
PCT for this break is 724 F.
Two HPCI line breaks were analyzed--one sufficiently large (.15 ft ) to 2
prevent any HPCI water from entering the vessel, and the other (0.045 ft )
such that the HPCI flow would approximate the break flow.
Smaller breaks would add HPCI water into the vessel, creating more fluid inventory and a more rapid depressurization rate from the subcooling of HPCI liquid.
Even without the emergency condenser, the 0.15 ft HPCI line break was shown to be non-limiting as the core remained covered except for the last 20 seconds before f
rated spray.
2 For the.045 ft HPCI line break, the emergency condenser was included in the analysis.
The emergency condenser was modeled with the condensation model described in Section 5.0.
This model predicted the design heat removal rate of the emergency condenser. The addition of condensation correlations i
{
caused the single volume secondary of the steam generator to unrealistically use a condensation correlation when a collapsed (completely liquid) pool f
covered the tubes.
The unrealistic behavior was corrected by dividing the secondary side of the steam generator into two vertical volumes, one for the bundle region and one for the steam dome region.
The results (Table 3.6) of the.045 ft break show that the core midplane was uncovered for 78 seconds prior to rated LPSI, with a resulting PCT of 1045 F.
Thus, the LPCS and HPCI line break analyses confirm that non-recirculation line breaks are non-limiting.
(
(
) XN-NF-79-95 (NP)
I Revision 1 l
I.
1 I,
i Table 3.1 Break Location Study, Based on j
1.0 DEG Break i
l Break Location Case
- 1 2
3 A PCT, F
-14 0
-104
- Case identification is as follows:
l
- 1) Pump suction piping, just upstream of pump.
j
- 2) Halfway between Pump and Steam Generator.
- 3) Cold leg, between Steam Generator and Reactor Pressure Vessel.
gl 1
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Table 3.2 Large Break Spectrum Event Time Results Time of Event After Break (soci Double-Ended Guillotine Breaks Split Breaks Break 2
2 2
2 2
Description 1.0 DEG 0.8 DEG 0.6 DEG 0.4 DEG 4.25 ft 3.40 ft 2.53 ft 1.69 ft 0.75 ft Steam Drum ECC Signal 6.1 6.1 6.1 6.5 6.8 6.3 6.1 6.6 10.1 Containment ECC Signal 3.0 3.1 3.2 3.5 3.2 3.2 3.2 3.6 5.2 Steam Drum Empty 13 13 13 13 13 13 13 13 16 HPCI Inj.*
20.0 20.1 20.2 20.5 20.2 20.2 20.2 20.6 22.2 47 LPCS Inj.*
60.4 58.7 59.7 65.3 8.5 58.4 58.5 63.4 94.4 Time of Rated Spray 67.6 68.4 69.1 76.4 67.5 67.5 67.8 74.2 114.3 Time of PCT 358 385 373 411 359 351 398 385 495 x5
- Start of coolant injection into vessel (pumps on and valves begin to open).
2 s.
77' 83 55a f
.)
Table 3.3 Large Break Heatup Results 3reak 2
2 2
2 2
Description 1.0 DEG 0.8 DEG 0.6 DEG 0.4 DEG 4.25 ft 3.40 ft 2.53 ft 1.69 ft 0.73 ft
% MWR 3.17 2.63 3.01 2.38 3.16 3.23 3.14 2.76 1.92 I
PCT OF 1924 1889 1914 1866 1925 1929 1923 1897 1823 Core Wide
<l.0
<l.0 cl.0
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<l.0
<l.0
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<1.0
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XN-NF-79-95 (NP)
Revision 1 Table 3.4 Small Creak Spectrum Event Time Results Time of Event After Break (sec) 2 2
2 2
Event 0.75 ft 0.50 ft 0.25 ft 0.10 ft Steam Drum ECC Initiation Signal 9.0 12.9 18.0 22.2 High Containment Pressure ECC Initia-5.3 7.1 14.2 250 tion Signal HPCI Injection
- 22.3 24.1 31.2 110.2 LPCS Injection
- 86.1 117.5 217.0 Time of Hot Plane Uncovery 55.0 57.0 60.0 Never uncovers Duration of Hot Plane Uncovery 45.0 11.0 11.0 0.0 Rated Core Spray 98.9 133.0 255.1 Time of PCT 450 422 0.2 0.2 l
- Start of coolant injection into vessel (pumps on and valves begin to open).
_ _ _ _ XN-NF-79-95 ( NP) l Revision 1 I
l Table 3.5 Small Break Heatup Results l
I 2
2 2
1 Event 0.75 ft2 0.50 ft 0.25 ft 0.10 ft l
PCT OF 1725 1404 702 702
% MWR
.92
.16
.06 Core Wide % MWR
<l.0
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I l
Table 3.6 Non-Recirculation Break Results 2
i
.045 ft j
Steam Drum low level ECC Initiation (sec) 26.8 22.2 j
Low Reactor Water Level ECC j
Inj. Permissive (sec) 735 201 HPCI Injection (sec) 735 201 l
Rated Core Spray (sec) 1985 Time of Hot Plane Uncovery (sec) 1907 315 f
Duration of Hot Plane U. covery (sec) 78 60
(
Time of PCT (sec) 2140 370 PCT ( F) 1045 724
% MWR 0.06
.06 Core Wide % MWR
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Corrected Page
)
4.0 HEATUP ANALYSIS RESULTS This section provides heatup analysis results for the application of ENC's WREM-Based NJP-BWR ECCS Evaluation Model to Dresden Unit 1.
The i
results support an ECCS allowable Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limit of 13.4 kw/ft for ENC fuel assemblies at begi nni ng-o f-l i fe. With exposure, the MAPLHGR limit decreases to values of 13.2 kw/ft at 5000 MWD /MTM,13.0 kw/ft over the range 10000 MWD /MTM to 27000 MWD /MTM, and 12.6 kw/ft at 33000 MWD /MTM.
For exposed Gulf United Nuclear assemblies with 35 active rods, the heatup analysis supports MAPLHGR limits of 13.2 kw/ft at 5000 MWD /MTM,13.0 kw/ft over the range 10000 MWD /MTM to 27000 MWD /MTM, and 12.6 kw/ft ac 33000 MWD /HTM. The l
MAPLHGR limits for exposed Gulf United Nuclear assemblies with 36 active j
rods are 12.6 kw/ft at 5000 MWD /MTM,12.4 kw/ft over the range 10000 MWD /MTM l
l to 27000 MWD /MTM, and 12.0 kw/ft at 33000 MWD /MTM.
The MAPLHGR results are 2
for the ENC limiting 3.40 ft split recirculation line pipe break.
The heatup analysis results are provided in Table 4.1.
Figure 4.1 provides the heatup transient for ENC fuel at beginning-of-life from the HUXY calculations.
Table 4.2 provides results of the axial power peak location sensitivity study.
In this study, the elevation of the limiting peak power node was assumed to vary from the nominal center peak location in both the RELAP4-EM/ HOT CHANNEL and HUXY/BULGEX calculations for the limiting exposure case. Axial power peaks at 2.0 ft below the top of the f
9.0 ft core and 2.0 ft above the bottom of the core were considered.
The 4
l
I 1
1 XN-NF-79-95(NP) 3 i
Revision 1 g
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l bottom peaked location yielded a slightly higher PCT and metal water reaction than the center peaked power profile.
The top peaked power profile had a l
j lower PCl and metal water reaction.
The calculations were made for the most limiting fuel type and exposure case in Table 4.1.
In summary, the resul ts given in Table 4.1, when combined with the axial study in Table 4.2, show that the calculated PCT and maximum local metal water reaction are less than j
the 10 CFR 50 allowable limits of 2200 F and 17%, respectively.
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TABLE 4.1 HEATUP ANALYSIS RESULTS SUMMARV Dresden Unit 1 6x6 Fuel Assembly Radial Peaking = 1.7 Planar Average Maximum Local MAPLHGR Exposure Metal Water Reaction Fuel Type (kw/ft)
(MWD /MTM)
PCT (*F)
(%)
ENC (35 Active Rods) 13.4 D
2024 4.6 13.2 5000 1992 7.3 13.0 10000 2035 12.1 13.0 15000 2091 14.8 13.0 21000 2053 13.1 13.0 27000 2049 12.8 12.6 33000 2031 12.1 h$
Gulf United Nuclear 13.2 5000 1946 3.3 13.0 10000 1996 10.9 (35 Active Rods) 13.0 15000 2034 12.5 13.0 21000 2030 12.4 13.0 27000 2023 12.0 12.6 33000 1987 10.5 Gulf United Nuclear 12.6 5000 1948 3.3 12.4 10000 1 974 9.9 (36 Active Rods) 12.4 15000 1993 10.8 12.4 21000 2013 11.5 12.4 27000 2005 11.1 12.0 33000 1960 8.2
-2 o~
CORE WIDE METAL WATER REACTION LESS THAN 1.0%
[, [x 4
TABLE 4.2 AXIAL POWER PEAK LOCATION SENSITIVITY STUDY j
f ENC FUEL, MAPLHGR = 13.0 kw/ft,15000 MWD /MTM j
l i
Axial Peak Location Maximum Local Above Bottom of Peak Clad Temperature Metal Water Reaction i
Active Core (ft)
( F)
(%)
2.0 (bottom skew) 2106 16.9 4.5 (center peak) 2091 14.8 1
7.0 (top skew) 2032 6.9 5' s i
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/ s3,.;,1
-S 121514 3 MAPLHGR 13.4 kw/ft
=
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.1 1 > 1. 20,' J i L'i ld P i.N c2, Ca ni s ter c,.)c y I
t 20?4 F
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!00 400 500 E 'O 7CC 300 9o
$1Y I b FIGURE 4.1 HEATUP TRANSIErlT - Ef1C FUEL - BOL XN-NF-79-95(NP)
Revision 1 5.0 MODEL HISTORY The following section presents the geneology of the RELAP4-EM f
used in the subject analysis.
5.1 RELAP4-EM GENE 0 LOGY RELAP4-EM/ ENC 28B - The code changes to RELAP4-EM/ ENC 26A to produce RELAP4-EM/ ENC 288 are described in the attachment to a letter to D. F. Ross from G. F. Owsley dated October 1978.
These code changes have since been approved as noted in Reference 7.
Code changes previously submitted to the NRC Staff (Reference 12) and which do affect calculated results are as follows:
{
RELAP4-EM/ ENC 28C - Three plot variables were added to RELAP4-EM/
ENC 288 to permit plotting of fuel related heat slab internal temperatures,
(
i.e., pellet surface temperature, clad inside surface temperature, etc.
A causal heat slab variable (time step control) was initialized to permit
('
execution of a RELAP4 case without a core and with zero heat slabs. These
{
changes do not affect calculated results.
RELAP4-EM/ ENC 28D - A change was made to output tape edit sub-f routine to allow output tape (TAPE 4) re-editing.
This change does not affect calculated results.
RELAP4-EM/ ENC 28E - The environmental package was modified to allow input data to be entered on the data cards in column 80.
This change does not affect calculated results.
f RELAP4-EM/ ENC 28F - The maximum number of words allowed on the input data cards was increased to permit larger problems. This change does not affect. calculated results.
d XN-NF-79-95(NP)
Revision 1 I
Recent code changes made to RELAP4-EM/ ENC 28 which have not been submitted to the NRC Staff and which do not change the calculated results are as follows:
RELAP4-EM/ ENC 28G - This was an experimental code version.
The I
coding introduced was removed in ENC 281.
l i
l RELAP4-EM/ ENC 28H - The number of time dependent volumes was 1
increased from 20 to 50.
The option of reading normalized power vs time El data in an Evaluation Model calculation was added.
These changes do not affect calculated results.
RELAP4-EM/ ENC 281 - The coding added in ENC 28G was taken out.
A variable in subroutine ENQAD that previously could be undefined was initial-ized to allow restarts.
These changes do not affect calculated results.
5.2 MODEL CHANGES FOR THE DRESDEN-1 ANALYSIS t
Model changes made to RELAP4-EM/ ENC 28 in tF: performance of the Dresden 1 ECCS analysis are discussed below:
RELAP4-EM/ ENC 28J - A more realistic method of calculating local heat flux in the vicinity of the core mixture level for small break ar.alysis has been developed.
The small break evaluation model previously used calculated heat flux from an adjacent core heat conductor for local heat flux on the covered region of the adjacent higher elevation conductor.
A detailed description of this model is on page 13 of Reference 8.
For long duration small break transients, covered regions of the core achieve a l
pseudo equilibrium condition such that the decay energy is transferred to i
I 4
I I
I i
XN-NF-7 9-95( NP)
Revision 1 the coolant at the decay heat generation rate.
For this condition, the covered region axial heat flux distribution will be directly related to the axial power distribution.
Thus, the small break model was modified so that the covered region heat flux is the adjacent conductor heat flux weighted by the ratio of the powers for the two conductors.
RELAP4-ENC 28K - This version improved the numerical iteration scheme for heat slab temperature initialization.
The modifications do not alter either the fuel / clad thermal expansion models or the gap con-ductance model.
Stored energy in the fuel is still determined from the GAPEX(II) model. A sample calculation for Dresden-1 confirmed a negligible impact of this change on PCT (<2 F).
RELAP4-EM/ ENC 28L - This version incorporated a turbulent con-densation correlation for use in the emergency condenser. When the surface temperature of a heat slab is below Tsat, Tbulk 1 Tsat, and the quality in the adjoining volume is greater than zero, then condensation l
is allowed as predicted by:
.5 y
f pf i'
/K)
.8
.4
'~
Re Pr 1+Xj#- -I l
= 0.023 l D )l
(
9
/_
hc
( h t
... _ _ ___. XN-NF-79-95(NP)
Revision 1 I
h where:
{
Btu condensation coefficient h
=
hr ft F
liquid thermal conductivity K
=
W ft F 7
hydraulic diameter (ft)
D
=
h liquid Reynolds Number Re
=
f liquid Prandtl Number Pr
=
f i
l volume average quality a
l X
=
i 3
pf liquid density (Ib /ft )
=
m p
vapor density (lb /ft )
=
g m
I This correlation was developed by Z. L. Miropolskiy, et al.(
for condensation of steam inside tubes.
The correlation is valid for all
\\
l anticipated blowdown pressures (60-3000 psia) and for mass fluxes of 37.9-379. (lb /ft sec). At lower mass fluxes, the correlation con-servatively underpredicts data as the effect of gravity in condensation films flows is not considered.
RELAP4-EM/ ENC 28M - Included in this version are added con-vergence criteria in the heat slab initialization subroutine.
These modifications are used only in the initialization.
A sample calculation for Dresden Unit 1 confirmed a negligible impact of RELAP4-EM/ ENC 28K and RELAP4-EM/ErlC28M modi fica tions on PCT (<2F ).
RELAP4-EM/ ENC 28N This version added a laminar condensation correlation for use in the emergency condenser.
Version [NC28L had I
1 l
I I'
XN-N F-79-95 (NP)
Revision 1 added a turbulent condensation correlation which extrapolated to unrealistically small values at low flows.
The condensation correlated added is the classical Nusselt(
} laminar condensation correlation for horizontal tubes as modified for the presence of residual condensate on the inside of the tubes.
~
I 3
2 g og g LH k
gg
_ pf D (T
-T )
sat g _
e where properties are evaluated at the average film temperature, and:
inside diameter of the tube D
=
enthalpy difference between saturated steam and AH
=
I9 subcooled liquid in the film factor which accounts for the reduction in heat F
=
transfer due to the residual liquid in the bottom of the tube.
F was devqloped from a least squares fit given in Collier (131,
- 77.la4 + 54.la5, 34,4a6 F =.043 + 3.37a - 18.2a2 + 53.0a3 a = void fraction The approved turbulent natural circulation heat transfer correlation used in RELAP4-EM FLOOD (") was incorporated into the blowdown heat transfer package. As discussed in Reference (15),
turbulent conditions are usually encountered in water systems.
For completeness, the laminar natural convection heat transfer correlation of Kreith, Reference (16) was included in this update.
l XN-NF-79-95(NP) i Revision 1 Corrected Page j
6.0 CONCLUSION
S A spectrum of breaks was analyzed in accordance with 10 CFR 50.46 and Appendix K for Dresden-1 and the results are presented in this document.
The limiting break was determined to be the 3.40 ft, large split break of 2
l a recirculation line occurring between the recirculation pump and the steam generator.
This analysis supports a Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) of 13.4 kw/ft for ENC 6x6 assemblies at begi nni ng-o f-l i fe. With exposure, the MAPLHGR limit decreases to values of 13.2 kw/ft at 5,000 MWD /MTM,13.0 kw/ft over the range 10,000 MWD /MTM to 27,000 MWD /MTM, and 12.6 kw/ft at 33,000 MWD /MTM.
For irradiated Gulf United Nuclear assemblies with 35 active fuel rods, the heatup analysis supports the same exposured MAPLHGR limits of 13.2 kw/ft at 5,000 MWD /MTM, 13.0 kw/f t over the range 10,000 tiWD/MTM to 27,000 MWD /t1TM,12.6 kw/ft at 33,000 MWD /MTM.
The MAPLHGR limits for irradiated Gulf United Nuclear assemblies with 36 active fuel rods are 12.6 kw/ft at 5,000 MWD /MTM,12.4 kw/ft over the range 10,000 MWD /MTM to 27,000 MWD /MTM, and 12.0 kw/ft at 33,000 MWD /MTM.
These limits are shown in Figure 1.1 Results of the present analysis (summarized in Figure 1.1 and Table 1.1) indicate that the Dresden-1 Emergency Core Cooling System (upgraded with the HPCI system), as analyzed with the ENC NJP-ECCS evaluation model, meets the acceptance criteria as presented in 10 CFR 50.46I4) and Appendix K.
That is:
(1) The calculated peak fuel element clad temperature does not exceed the 2200 F limit.
i
6 XN-NF-79-95(NP)
Revision 1 I
(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 when the core geometry is still amenable to cooling.
The hot I
fuel rod cladding oxidation limits of 17% are not exceeded during or after quenching.
1 I
I i
I I
i I
2 XN-NF-79-95 (NP)
Revision 1
7.0 REFERENCES
1.
The Exxon Nuclear Company WREM-based NJP-BWR ECCS Evaluati'on Model and Application to the Oyster Creek Plant, XN-75-55, Revision 2, XN-75-55, Revision 2, Supplement 1, XN-75-55, Revision 2, Supplement 2.
2.
Safety Evaluation Report by the Office of Nuclear Reactor Regulation Regarding Review of the Exxon Nuclear Company Non-Jet Pump Boiling Water Reactor ECCS Evaluation Model Described in Exxon Topical Reports XN-75-55, Revision 2, dated August, 1976, XN-75-55, Revision 2, Supplement 1, dated September,1976.
XN-75-55, Revision 2, Supplement 2, dated December,1976, for Conformance to Appendix K to 10 CFR 50, USNRC, February 25, 1977.
3.
ECCS Evaluation of Dresden-I using the Exxon Nuclear Company WREM Non-Jet Pump Evaluation Model - Large Break Example Problem, XN-NF-79-24, March, 1979.
4.
U.S.N.R.C., Acceptance Criteria for Emergency Core Cooling System for Light Water-Cooled Nuclear Power Plants,10 CFR Part 50, Federal Register, Vol. 39, No. 3, January 4,1974.
5.
Exxon Nuclear Company, HUXY: A Generalized Multirod Heatup Code with 10 CFR 50 Appendix K Heatup Option Users Manual.
XN-CC-33(A),
Revision (1), November,1975.
6.
Big Rock Point LOCA Analysis Using the ENC NJP-BWR ECCS Evaluation Model, XN-NF-78-53, December,1978.
I 7.
U.S.N.R.C. letter, T. A. Ippolito (NRC) to W. S. Nechodom (ENC),
SER for ENC RELAP4-EM-UPDATE, March, 1979.
l 8.
Exxon Nuclear Company WREM-based Generic PWR ECCS Evaluation Model, XN-NF-75-41, Volume III, Revision 2, August 20, 1975.
i
- 9. - Miropolskiy, Z.
L., Shneerova, R.
I., and Ternakova, L. M.:
Volume III, Int. Heat Transfer Conf., Japan, Cs. 1.7 (1974).
I 10.
BULGEX: A Computer Code to Determine the Deformation and the Onset of Bulging of Zircaloy Fuel Rod Cladding, XH-74-27, Revision 2, i
December. 31, 1974.
11.
Exxon Nuclear Company, GAPEX: A Computer Program for Predicting Pellet-to-Cladding Heat Transfe-Coefficients, XN-73-25, August 13, 1973.
1 XN-NF-79-95(NP)
Revision 1 I
l 12.
Exxon Nuclear Company, Fort Calhoun LOCA Analyses at 1
1500 MWt Using ENC WREM-II A PWR ECCS Evaluation Model, XN-NF-79-89, September,1979.
13.
Convective Boiling and Condensation, John G. Collier, McGraw-Hill Book Co., 1972.
- 14. WREM Water Reactor Evaluation Model, NUREG-75/056, Rev.1, May 1975, pp. 5-14 1
15.
Brown, A.I. and Marco, S. M., Introduction to Heat Trans fer, McGraw-Hill Book Co., Third Edition,1958, page 165.
16.
Kreith, F., Principles of Heat Transfer, International Textbook Company, Fourth Printing,1951, page 306.
I!
l l!
I Il I'
I f
I, I
I!
I I
XN-NF-79-95(NP)
Revision 1 04/24/80 ECCS EVALUATION OF DRESDEN-1 USING THE EXXON NUCLEAR COMPANY WREM-BASED NON-JET PUMP EVALUATION MODEL Distribution DJ Braun RE Collingham GC Cooke RD Hyman SE Jensen DC Koletar JE Krajicek CD May JN Morgan GF Owsley FB Skogen GA Sofer PJ Valentine CECO /CD May (55)
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