ML20086K511
| ML20086K511 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 10/31/1991 |
| From: | Braun D SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML19353B429 | List: |
| References | |
| EMF-91-172, NUDOCS 9112130147 | |
| Download: ML20086K511 (48) | |
Text
__
I SIEMENS I
EMF 91-172 I
I I
l Grand Gulf Unit 1 LOCA Analysis for Single Loop Operation i
I I
I I
October 1991 1
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1b 10. ' tJ'3i) r 4
SIEMENS e,,.,,.,,,
Issuo Dato:10/ll/91 GRAND GULF UNIT 1 LOCA ANALYSIS FOR SINGLE LOOP OPERATION Prepared by:
J B-7J. Braun BWR Fuel Engineering Fuel Engineering and Ucensing October 1991
/smg Siemens Nuclear Power Corporation Eng nee < ng anc Manufactu' ng Fa 4!/
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<509i3'i5C2 2!01 Horn Aao'as Acac PO Bov 133 R craano WA 99352 0t30 Te* t509, 375 6100 a
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CUSTOMER DISCt. AIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFyj.R Semens Nuc6eer Power Carperston's warrantes and representatons concemog me sutgeot masser of tres document are those set for1n m the Agreement be, ween Seement Nuoleer Power Corporoton and me Customer pursuant to wrwch es docunent e issued Accon$ngly, esoept as omerwise expressay proveed m such Agreement, nemer Somens Nucteer Power Corporabon not any person actng on its behalf makes any warfanty or representsaan orpressed or imphed, with respect to the accuracy, compioennees. or usefulness of me mtormason contamed m he document, or that me une of any informaeon apparatJs, asethod or process
-in me document wd not minnge privately owned nghts: or assumes any habeless wei respect to me Joe of any infomsabon. apparaNs, method or process W in the document.
The informaton contaned horen is for the sono use of the Customer.
in order a svoed impeemmt of nghts of Semons Nucioar Power Corporauon e pasents or ewoneens whmh may be induced m the mformaton contained m es domment, me recipsort, by as accopsence of his document, agrees not to pubbsh or make putec use (m me pesant use of the term) of such mformaton unal so aumonzod in wrong by Siemens Numeer Power Corporaton or enol after sur (6) months fosowmg isrmnason or expeenon of the aforesed Agnumma and any eriewan thereef. unises espressly provided in the Agreement. No ngnts or hconses e or to any passnes are enphed by me fumierung of ha document
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I EMF.91 172 Pagei i
i TABLE OF CONTENTS 1
1 Section Pace
1.0 INTRODUCTION
1 2.0
SUMMARY
2 3.0 JET PUMP BWR ECCS EVALUATION MODEL........
3 3.1 LOCA During Single Loop Operation..........
3 3.2 EXEM/BWR Application To Grand Gulf Unit 1 4
4.0 ANALYSIS RESULTS AND CONCLUSIONS..........................
8
5.0 REFERENCES
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EMF.91 172 Page ii UST OF TABLg1
.T_a.ple E.aan 3.1 GRAND GULF UNIT 1 REACT OPERATION.............OR SYSTEM DATA FOR SINGLE LO 4.1 LOCA EVENT TIMES SINGLE LOOP OPERATION........
6 4.2 PCT RESULTS FOR SNP FUEL SINGLE AND TWO LOOP O 10 11
- UST OF FIGURES fl2Et Pat 3.1 GRAND GULF UNIT 1 SYSTEM BLOWDOWN NODALIZAT 4.1 UPPER PLENUM PRESSURE........
7 4.2 TOTA L B R EA K FL OW.............................................
12 4.3 AVERAGE CORE INLET FLOW....................................
13 4.4 AVERAGE CORE OUTLET FLOW.................................
14 4.5 HOT CHANNEL INLET FLOW...................................
15 4.6 HOT CHANNEL MIDPLANE FLOW..........
16 4.7 HOT CHANNEL OUTLET FLOW...................................
17 4.8 INTACT LOOP JET PUMP DRIVE FLOW........................
18 4.9 INTACT LOOP JET PUMP SUCTION FLOW..........
19 4.10 INTACT LOOP JET PUMP EXIT FLOW..........................
20 4.11 BROKEN LOOP JET PUMP DRIVE FLOW....................
21 4'12 ~ BROKEN LOOP JET PUMP SUCTION FLOW.........22 4.13 BROKEN LOOP JET PUMP EXIT FLOW........................
23 4.14 LOW PRESSURE COOLANT INJECTION FLOW........
.................24 4.15 HIGH PRESS URE CORE SPRAY FLOW........................
25 4.16 LOW PRESSURE CORE SPRAY FLOW............
................. 26-4.17 UPPER DOWNCOMER UQUID MASS.............................
27 4.18 MIDDLE DOWNCOMER UQUID MASS..........................
28 4.19 LOWER DOWNCOMER UQUID MASS.................
............... 29 4.20 LOWER DOWNCOMER MIXTURE LEVEL.,.
30 4.21 UPPER PLENUM UQUID MASS.............
31 4 22 LOWER PLENUM UQUID MASS................................
32 4.23 REFILUREFLOOD SYSTEM PRESSURE..........................
33 4.24 REFILUREFLOOD LOWER PLENUM MIXTURE LEVEL,................
34 4.25 REFILUREFLOOD RELAllVE CORE MIDPLANE ENTRAINMENT 35 4.26 HOT CHANNEL CENTER SLAB HEAT TRANSFER COEFFICIENT 36 4.27
. HOT CHANNEL CENTER VOLUME QUAUTY,.....,.................
37 4.28 HOT CHANNEL CENTER VOLUME COOLANT TEMPERATURE....,...
38 4.29 HOT ASSEMBLY HEATUP RESULTS FOR SNP 9x9-5 FUEL (BO 39 40 i
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EMF 91172 Page 1
1.0 INTRODUCTION
The results of a loss-of coolant accident (LOCA) analysis with emergency core cooling systems (ECCS) for the Grand Gulf Unit 1 to suppcrt single loop operation (SLO) are reported in this document. These calculations are performed with the generically approved Siemens Nuclear Power Corporation (SNP) EXEM/BWR Evaluation Model ) in accordance with U
Appendix K of 10 CFR 50,M and the results comply with the USNRC 10 CFR 50.46(8) criteria.
The initial SLO condition selected for this analysis is 70.6% power and 54.1% flow. This analysis establishes the multiplier that is to tn applied to the two-loop MAPLHGRs of the fuel during single loop operation. The analysis is performed with the same multiplier for both SNP 8x8 and 9x9-5 fuel.
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EMF 91 172 Page 2 2.0 SUMMAFlY The results of the LOCA-ECCS analysis presented herein support the use of a 0.86 SLO multiplier on the tva-loop MAPLHGRs for SNP fuel when the Grand Gulf Unit 1 reactor is in single loop operation.
Single loop operation of Grand Gu!f Unit 1 with a multiplier of 0.86 on the two loop SNP fuel MAPLHGRs for 8x8 and 9x9-5 fuels assures that the emergency core cooling systems for the W or loss of-Grand Gulf Unit 1 plant will meet the USNRC acceptance criteria of 10 CFR 50.46 f
coolant accident breaks up to and including the double ended severance of a reactor coolant pipe. That is:
1.
The calculated peak fuel element clad temperature does not exceed the 2200*F limit.
2.
T he calculated total oxidation of the cladding nowhere exceeds 17% of the total cladding thickness before oxidation.
3.
The calculated core wide metal-water reaction does not exceed 1% of the zircaloy associated with the active fuel cladding in the reactor.
4.
The LOCA ciadding temperature transient is calculated to be terminated at a time when the core is still amenable to cooling.
5.
The system long-term cooling capabilities provided for the initial core and subsequent reloads remain applicable to SNP fuel.
1 l
EMF 91 172 Page 3 3.0 JET PUMP BWR ECCS EVALUATION MODEL 3.1 LOCA Durino sinole Loco Operation The loss-of-coolant accident (LOCA) break spectrum analysis for a BWR/G for two. loop operation conditions is described in Reference 7, and the Grand Gulf Unit 1 plant specific MAPLHGR analysis for two loop operation is described in Reference 8. This document describes LOCA-ECCS analysis and MAPLHGR justification for Grand Gulf Unit 1 SLO operation. The same limiting break and single failure are assumed in both the two loop and the SLO LOCA analyses.
During SLO the recirculation pump in the inactive loop is not in operation and the recirculation flow contiof valve is put at minimum position. Both intake and discharge block valves in the idle recirculation loop remain open during SLO. A significant resistance to idle loop recirculation flow does exist, but a small amount of flow can pass through the idle loop during LOCA conditions. A break can be hypothesized to occur in either loop. However, a break in the inactive loop would behave essentially like a break during two loop operation except that substantial break flow would come from only one side of the break (because the recirculation flow control valve is at its minimum position). System performance would then be like that resulting from a somewhat smaller break during two loop operation. The scenario of breaks smaller than the limiting break has already been covered by the BWR/6 LOCA break spectrum analysis.M Further consideration in this report will be given only to the case where a break occurs in the active loop. This case differs from the two loop case in ono important respect:
there is no flow coastdown in the intact (idle) recirculation loop.
Previous SLO analysis 00) assumed that the consequence of a lack of recirculation loop coastdown flow (which continues to supply liquid from the downcomer to the lower plenum during two loop operation) would be an almost immediate flow stagnation in the core and a very early CHF (0.1 sec); this resutted in degraded heat transfer very early in the transient and 09 be imposed for SLO conditions on the required that a MAPLHGR reduction factor of 0.86 MAPLHGR for NSSS vendor fuel for Grand Gulf Unit 1.
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EMF 41 172 page 4 3.2 EXEM/BWR Aeolication To Grand Gutf Unit 1 The SNP EXEM/BWR ECCS Evaluation model codes are used for this SLO LOCA-calculation.
The codes which comprise EXEM consist of RELAX.(2) FLEX,W HUXY,FS)
RODEX2.W The latest versions of these codes are used for this SLO analysis, and they are referenced in the two loop analyses?'8)
The approved SNP generic break spectrum analysisM for the BWR/6 class of reactors identified the limiting break as the double-ended guillotine break of the recirculation pump discharge pipe with a discharge coefficient of 1.0 (1.0 DEG/RD) and a worst single failure.
In the unlikely event that a LOCA would occur during SLO, a rapid drop in core flow would be expected to occur during the early phase of the event because the idle loop pump is not operating. The core flow transient during the early phase (0 to 5 seconds) of a single loop LOCA is the principal event which could distinguish such an accident frcm a LOCA occurring during normal two loop operation. Other than the core flow, the LOCA results for two and single loop operations are quite similar, as shown by event times (time of uncovery of jet pumps, lower plenum flashing, time of rated spray, etc.). Therefore the worst single failure, the size, location, and discharge coefficient of the break for the single loop LOCA will be the same as for the two loop LOCA.
The 70.6% power /54.1% core flow (70.6/54.1) operating point is selected as the initial operating condition for this analysis. The conditions are those used by the NSSS vendor in their SLO LOCA-ECCS analysis.(11)
The system behavior during a LOCA is determined primarily by the LOCA break parameters: break location, break size, and break configuration together with the ECCS systems and plant geometry. Variation in core geometric parameters produce only secondary effects on the system behavior. Thus, 'by using bounding core neutronic parameters, the LOCA ECCS results established by this analysis will apply for future cycles unless substantial changes are made in the plant operating conditions, plant hardware, or core design such that the analysis no longer bounds the plant conditions.
EMF 91 172 Page 5 The system blowdown calculations for the Grand Gulf Unit 1 SLO differ from those for two loop operation in that the back flow through the idle loop jet pump is modeled consistent with expected SLC steady state conditions. The initial conditions are shown in Table 3.1, and the system blowdown nodalization diagram is shown in Figure 3.1.
The system blowdown calculations are followed by HOT CHANNEL calculations. The hot channel geometry is modeled to that of SNP 8x8 and 9x9-5 fuel residing in the Grand Gulf reactor. The initial conditions are adjusted to correspond to those for the single loop operating point. The power, axial peaking, and flow of the hot assembly are determined by an XCOBRA thermal hydraulics calculation to support a MCPR value considerably below the appropriate MCPR operating limit. The nodalization and geometry used in the raflood calculation are
. identical to those of the two loop analysis. In the FLEX code the intact loop is not modeled in detail because intact loop flows are insignificant at the time of rated low pressure core spray.
Thus, no changes are required in the FLEX nodalization or geometry compared to the two loop analysis. The initial conditions for the reflood calculation are entirely determined by the system blowdown calculation.
The HUXY heatup calculation of the hot plane (conter one foot node)is done identically with previous two loop analyses: fuel stored energy, gap thermal conductivity and dimensions from RODEX2 as a function of power and exposure; time of rated spray, decay power, heat transfer coefficients and coolant conditions from RELAX; and time of hot-node-reflood from FLEX.
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 heatup calculation is conservative. The spray heat transfer coefficients identified in Appendix K of to CFR 50(8) are used for the SLO analysis in an identical manner as used in the approved two loop analyses (sa2),
This includes the use of 5.0 Btu /hr ft,.p 2) for all unheated surfaces in the SNP 9x9-5 fuel. Peak 2 p cladding temperature (PCT) and the cladding oxidation percentage are specifically determined
- for the SNP fuel geometries.
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EMF 91 172 Page 6 TABLE 3.1 GRAND GULF UNIT 1 REACTOR SYSTEM DATA FOR SINGLE LOOP OPERATION Primary Heat Output, MW (1.02 x.706 x 3833) = 2760.2 8
8 Total Reactor Flow Rate, Ib/hr
(.541 x 112.5 x 10 ) = 60.86 x 10 s
Active Core Flow Rate, Ib/hr
. 54.28 x lo Steam Dome Pressure, psia 1001.7 Reactor Inlet Enthalpy, Btu /lb 511.8 6
Recirculation Loop Flow Rate,Ib/hr 15.93 x 10 8
Steam Flow Rate,Ib/sec 11.47 x 10 6
Feedwater Flow Rats, Ib/sec 11.44 x 10 Rated Recirculation Pump Head, ft 765.
' Rated Recirculation Pump Speed, rpm 1785.
2 Moment of inertia, Ibm-ft / rad 18,875 Recirculation Suction Pipe 1.0., in 21.825 Recirculation Discharge Pipe I.D., in 21.825 4
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EMF 91 172 Page8 4.0 ANALYSIS RESULTS AND CONCLUSIONS The results are obtained by a LOCA.ECCS heatup analysis of the 1.0 DEG/RD break over l
the exposure range of the SNP 8x8 and 9x9-5 fuels in the Grand Gulf reactor. The fuel stored i
energies at 5 GWd/MTU are used to initialize the RELAX / HOT CHANNEL calculations for both fuel j
types. These stored energies are the maximum calculated by the RODEX2 code to occur during the life of the fuels. This means that the fuel temperatures in the hot channels are at their maxirm m calculated values regardless of the fuel exposure at which the pipe rupture is assumed to occur.
Shown in Table 4.2 are the highest PCTs over the exposure range for both fuels in both single and two(813) loop operation. The two loop MAPLHGRs with a.86 multiplier used in this single loop analysis produce lower PCTs than the acceptance criteria of 2200 'F and are less than the PCTs for the two-loop analysis by about 100 'F. Thus, a.86 multiplier on the two loop SNP 8x8 and 9x9-5 MAPLHGR curves assures that the 10 CFR 50.46(8) criteria are met under SLO conditlans.
Table 4.1 lists the major event times for this analysis. Syster "'owdown results are given in Figures 4.1 through 4.22. System refill and reflood results are si own in Figures 4.23 through
'4.25. Results from the SNP 9x9-5 RELAX / HOT CHANNEL calculation are given in Figures 4.26 through 4.28. The RELAX / HOT CHANNEL results are applied as boundary conditions for the HUXY heatup calculations. A clad temperature history determined by HUXY for the SNP 9x9-5 fuel at the beginning-of-life (BOL) exposure is shown in Figure 4.29.
An examination of these plots reveals the following information:
1.
The sudden loss of drive flow in the operating (broken loop) jet pumps results in a sudden drop in lower clenum pressure of sufficient magnitude to allow flow through the inactive jet pumps to " reverse" from their initial negative flow to a positive flow at one secqnd into the blowdown.
EMF 91172 Page 9 2.
The exit junction flows of the operating jet pumps reverse direction at one second into the blowdown.
3.
Bxause of 2. above, the initial drop in core flow is sufficient to cause an early CHF at the mid-plane of the SNP 8x8 and 9x9-5 hot channels. CHF occurs at less than 2 seconds for the SNP 8x8 and 9x9-5 fuel types.
With the results presented here, it it concluded that the application of a.86 multiplier during SLO conditions to the SNP 8x0 ar.d 9r9-5 fuel MAPLHGRs in the Technical Specifications will protect SNP fuelin the Grand Gulf Unit 1 ceactor from exceeding 10 CFR 50.4GN limits.
I
EMF 91 172 Page 10 TABLE 4.1 LOCA EVENT TIMES SINGLE LOOP OPERATION gvgn!
Time (sec)
Start 0.00 Initiate Break 0.05 Feedwater Flow Stops 3.05 Steam Flow Stops 6.05 Low Mixture Level for HPCS 10.2 Low MixNro Level for LPCS/LPCI 17.2 i
Jet Pumps Uncover 19.5 HPCS Flow Starts 20.2 Lower Plenum Flashes (Quality > 0) 21.3 Recirculation Suction Nozzle Uncovers 31.1 LPCS/LPCI Pressure Permissive Cleared 67.5 LPCS Flow Starts 82.7 LPCI Flow Injection Starts 91.5 End of Blowdown (Upper Plenum Pressure for 113.6 j
Raied LPCS flow)
Depressurization Ends (vessel pressure 180.4 l
reaches 1r.tm)
Start of Reflood (high density fluid 183.9 enters core)
Peak Clad Temperature Reached 194.5
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EMF 91 172 Page 11 TABLE 4.2 PCT RESULTS FOR SNP FUEL SINGLE AND TWO LOOP OPERATION Maximum PCT ('F)
Maximum PCT ('F)
.86 x MAPLHGR Two Looo Operation SLO SNP 8x8 Fuel 1738W 1631 SNP 9x9-5 Fuel 171303) 1609 a
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e EMF-91-172 Page 12 P
GRfhD GULF UNIT i Sf: 'r.0 72.012%P/54.1%F 1.0 DC3MD 8x8 FUEL
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EMF 91 172 Pago 13 GRAND Gulf UNIT 1 RC SLO 72.012%P/54.1%F 1.0 DEG/RD 8X8 FUEL i
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EMF 91 172 Page 15 GRAND GULF UNIT 1 AC SLO 72.012XP/54.1%F 1.0 0CG/RD 8X8 FUCL g
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GRAND GULF-UNIT 1 HC SLO 72.012'/.P/54.1%F 1.0 DEG/R0 9X9-5 FUEL-i i
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EMF-91 172 Page 17 GRAND GULF UNIT 1 HC SLO 72.012%P/54.1%F 1.0 DEG/R0 9X9-5 FUCL i
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FIGURE 4.6 HOT CHANNEL MIDPLANE FLOW
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EMF-91-172 Page 18 GRAND GULF UNIT 1 hC SLO 72.012%P/54.17.F 1.0 OCG/RD 9X9-5 FUCL i
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EMF 91 172 Page 19 GRANO GULF UNIT 1 AC SLO 72.012%P/54.1%F 1.0 DEG/RD 8X8 FUCL g
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EMF-91 172 Page 20 GRAND GULF UNIT 1 AC SLO 72.012*/.P/54.1%F 1.0 DEG/R0 8x8 FUEL
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FIGURE 4.9 INTACT LOOP JET PUMP SUCTION FLOW
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EMF 91 172 Page 21 GRAND GULF UNIT 1 AC SLO 72.012*/.P/54.1*/.T 1.0 DEG/R0 8X8 FUEL el i
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4 EMF 91-172 Page 22 GRAND GULF UNIT I RC SLO 72. 012%P/54.1%F 1.0 OCG/R0 8X8 FUEL I
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EMF-91-172 Page 23 8
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' EMF-91 172 Page 24 g
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EMF 91 172 Page 25 GRAND GULF UNIT 1 RC SLO 72.012%P/54.l*.F 1.0 DCG/R0 8X8 FUCL
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EMF-91 172 Page 26 GRAND GULT UNIT 1 AC SLO 72.012%P/54.1%F 1.0 DEG/RD 8XS FUEL i
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FIGURE 4.15 HIGH PRESSURE CORE SPRAY FLOW E
EMF-91-172 Page 27 GRAND GULF UNIT I AC SLO 72.012%P/54.1%F 1.0 DEG/R0 8x8 FUCL i
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EMF 91 172 Page 29 R
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FIGURE 4.10 MIDDLE DOWNCOMER t.;OUID MASS
EMF-91-172 Page 30 g
GRRNO GULF UNIT 1 AC SLO 72.012%P/54.1%r 1.0 DCG/R0 8x8 FUCL I
i l
3 4
I i
e_l
$h
~
e B
3l
=
i 5
8ag f
f I
f f
f 9
c0 20 40 60 80 100 120 140 160 TIMF ISCC)
FIGURE 4.19 LOWER DOWNCOMER UOUID MASS l
l i
l
EMF 91172 Page 31 GRAND GULF UNIT I AC SLO 72.0127.P/54.1%F l.0 OCG/R0 8x8 FUCL i
i i
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gz d
s d
b.
a -.
LJ b0 22 9s 8"
.=
m.
r t
i 1
I I
I g
a 40 60 80 100 120 140 160 T tic (SEC)
FIGURE 4.20 LOWER DOWNCOMER MIXTURE LEVEL -
e
o EMF-91 172 Page 32 GRAND GULF UNIT 1 AC SLO 72.012'.P/54.1%F 1.0 DEG/R0 8X8 FUCL
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i i
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e 8R JR I
L
~
~
5 L
bl 1
t t
t f
1 I
cc 20 40 60 80 100 120 110 160 T!?1C (SCC)
FIGURE 4.21 UPPER PLENUM LIQUID MASS
!j.
I i
EMF 91 172 Page 33 GRAND GULF UNIT 1 AC SLO 72.012%P/54.1%F 1.0 DEG/RD 8x8 FUCL i
i i
a i
s i
sl-s.
81
~3 -
I rg m
m'
.N' S
~
e i
i i
i r
a m
10 60 80 100 120 IM 160 TINC'(SEC1 FIGURE 4.22 LOWER PLENUM LIQUID MASS
EMF 91 172 Page 34 1
15 0 12$-
- ?
' 10 0-Y 8
75-g E
1Wp so-m 25-no 12e - tio tio 15 0 150 do
'50 tio 20o zio 22o 23o
- TiuE (SEC)
FIGURE 4.23 REFILUREFLOOD SYSTEM PRE'SSURE
/
e
- - - - - - - - - - - - - - - - - - - - - - - " - - - - - - - - - - - - - - ~ - - - - - - " ' - - - ~ - - ' - - - - - - ~ ~ - - ' - ' '
1 EMF 91 172 Page 35 25 p 2o-b, d
b e
w 15 -
o:
3 252 3
$ 10 -
-wd.
5m 3' 3 0
11 0 12 0 l$0 34 0 150 15 0 do t$0 15 0 250 2io 220
'30 TIME (SEC).
FIGURE 4.24 REFILUREFLOOD LOWER PLENUM MIXTURE LEVEL e
.,u s
g...
EMF-91 172 Page 36-
'I l
'1.25 2w w...
6i
)
d 3
0.S0 -
i
$p O.25-r...
_ 4.
TIME (SEC)
FIGURE 4.25 REFILUREFLOOD REl ATIVE CORE MIDPl.ANE ENTRAINMENT d
y-rc
..,-#+m.m
s I'
EMF-91 172 Page 37 GRAND GULF UNIT 1 HC SLO 72.012%P/54.1%f 1.0 DEG/R0 9X9-5 FUEL tfl i
I 4
3 4
=
- l
=r.
=
b.=
m b
a 8
E
[?.
d 5
E d
h E.
s
-1 h
a t
t I
t t
i 1
~
=
2 20 40 60 80 100 120 110 160 TINC (SCC)
FIGURE 4.26 HOT CHANNEL CENTER SLAB HEAT TRANSFER COEFFICIENT
pg, 4.
S L
EMF-91 172 Page 38 GRAND GULF UNIT 1 HC SLO 72.012'/.P/54.l*.F 1.0 DEG/R0 9x9-5 FUEL
/
es l
c-c..
g O
5 Ed i
Me
~
I 1
1 t
i t
cio 20 40 60 80 100 120 140 160 Ilr0 tSEC)
FIGURE 4.27 HOT CHANNEL CENTER VOLUME QUAUTY
e EMF-91 172 Page 39 GRAND GULF UNIT 1 HC SLO 72.012%P/54.1%F 1.0 OCG/R0 9X9-5 FUCL
.gf i
i i
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CH e
8_
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~
~
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3
('
s li ls
~
e o
k O
a u
Mg g
i I
t I
f I
-Eb 20
.J A
to 100 120 140 iso Titic (SCC)
FIGURE 4.28 HOT CHANNEL CENTER VOLUME COOLANT TEMPERATURE
.t I
EMF 91 172 Page 40 GRAND GULF 1 HUXY SLO 72.012*.P/54.l'$ BOL 1.0CG/R0- 9X9-5 FUCL
/
I'.2 3 4 56789 2 laIl1213itil1817 3 111819 20 al 22 23 24 4 1219252527292930 5 13 20 M 3132 33 34 35 4 it 2127 32 38 3? 3e 30 7 is 22 2s 33 sr,o u o 8 le 23 29 34 30 ti t3 44 1
- 17 74 30 35 M 42 +4 95 r
^
PIN 27 PIN 5 C 2000 C
CMISTER l.
6 '.
t
- 1500 n
4 O-lk Soo
_&e ww wa w
- w w e
t 1
1 1
f a
g g
g 30 60 90 120 150 100 210 240 f!PC 15CC) l'
$0110 OS/13/91 t-FIGURE 4.29 HOT ASSEMBLY HEATUP RESULTS FOR SNP 9x9-5 FUEL (BCu) i l
r P
o
- ' u-'.
- ~.., _. - -., _.., _,, _, _ _. -
4 EMF 91172 Page 41
5.0 REFERENCES
1.
- Exxon Nucle
&w
'3y for Boiling Water Reactors: EXEM ECCS Evaluation Model.
Summa /y Des, on."
NF-8019(A), Volume 2. Revision 1 June 1981.
2.
- RELAX: A RELAm wased Computer Code for Calculating Blowdown Phenomena,"
XN NF 8019(A), Volume 2A, Revision 1, Exxon Nuclear Company, June 1981, 3.
- FLEX: A Computer Code for the Reful and Reficod Period of a LOCA," XN NF 8019(A), _
Volume 2B, Revision 1. Exxon Nuclear Company, June 1981.
4.
"HUXY: A Generalized Multirod Heatup Code with 10 CFR 50 Appendix K Heatup Option User's Manual," XN CC-33(A), Revision 1 Exxon Nuclear Company, November 1975.
5
" Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model,"XN NF-82 07(A),
January 1982.
6.
"RODEX2: Fuel Rod Thermal Mechanical Response Evaluation Model," XN NF 8158(A),
Revision 2, Exxon Nuclear Company, March 1984.
7.
" Generic LOCA Break Spectrum Analysis for BWR/6 Plants," XN NF 86 37(P), Exxon Nuclear Company, Inc., April 1986.
8.
- Grand Gulf Unit 1 LOCA Analysis," XN-NF 86-38, Exxon Nuclear Company, Inc.,
June 1986.
9,.
" Acceptance Criteria for Emergency Core Cooling Systems for Ught Water Cooled Nuclear Power Reactors," 10 CFR 50.46, and "ECCS Evaluation Models," Appendix K of 10 CFR 50.
10.
" General Electric Company Analytical Model for Loss-of Coolant Analysis in Accordance With 10 CFR 50 Appenc x K, Amendment 2 One Recirculation Loop Out of Service,"
NEDO 20566-2, Revision 1. Class 1, July 1978.
11.
" Grand Gulf Nuclear Station Updated Final Safety Analysis Report," Appendix 15C.
12.
" Grand Gulf Unit 1 Cycle 5 Reload Analysis," ANF 90-022, Revision 2, Advanced Nuclear Fuels Corporation," August 1990, 13.
" Grand Gulf Unit 1 Citle 6 Reload Analysis" EMF 91169, Siamens Nuclear Power Corporation," October 1991.
4 EMF-91 172 Issue Date:10/11/91 GRAND GULF UNIT 1 LOCA ANALYSIS FOR SINGLE LOOP OPERATION Distribution D. J. Braun R. A. Copeland L J. Federico O. E. Garber N. L Garner D. E. Hershberger M. J. Hibbard J. N. Morgan R. S. Reynolds C, C. Roberts, Jr.
A. W. Will Entergy Operat!ons/S. L Leonard (40)
Document Control (5)
's d
b
l
- to GNRO 91/00186 1
l l.
l 1
-