ML20207G773
| ML20207G773 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 05/31/1986 |
| From: | Braun D, Ingham J, Krajicek J SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML20207G733 | List: |
| References | |
| TAC-61930, XN-NF-86-38, NUDOCS 8607230165 | |
| Download: ML20207G773 (53) | |
Text
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Xh-h =-86-38 I -
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.t GRAN J GU_F UNIT i,LOCA Ah A_YSIS I
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4 l- N AY 1986 1
L RICHLANJ, WA 99352
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E EXXON NUCLEAR COMPANY, INC.
388728a8 88800:16 p PDR l
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XN-NF-86-38 -
Issue Date: 5/30/86 f
GRAND GULF UNIT 1 LOCA ANALYSIS a
i Prepared by: C/l,f d l J. G. 2ngWam BWR Safety Analysis i
Prepared by: '
- g. E. @ ajicek, Sr. Engineer BWR Safety nalysis Prepared by: .
A / S.
l C D. // Braun'/
BWR 3afety Analysis
. Concur:
l.h' R.' E'.
f$
C511in fam, Manager
, BWR Safety A lysis I
3 Concur: hd .A Ab Qil/AJ rohL2 G. N! Ward,4Mandger '/
- Reload Licensing Concur: Y J. N. Morg SWett Manager b k{ /
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/
Customer Se ices g eering Approve: [ A f//rM/
[H.E.Wil son, Knager Licensing Safety Engineering i Approve: -
0 4 i
G. L. Ritter, Manager Fuel Engineering & Technical Services i
- tmrc ERON \UC_ EAR COV 3A\Y \C.
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CUSTOMER DISCLAIMER IMpORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT l PLEASE READ CAREFULLY ,
l Eamon Musiser Company's smarrendes and representmoons concoming es edigest metter of this document are those est fore in to Ayeoment hoeween Eamon N& Comsany, Inc. and the Cuesomer pursuant to whis ?
f this donament is issued. _ .4,, esempt a otherunse assueenly provided in such ^, ., nenhor Eamon Nuolour Company, ins, nor any person audng on its behalf makes any warretty or representadon, expressed or g
implied, whh respect to me assuracy, compisseness, or usefulness of the I information contemed in this desument, or that es use of any informenon, esperatus, neethod or process discioned in this document will not infrings prevensly ommed rights; or assumes any liabilities wie respect to the use of any :i.. z , apperseus, monod or prosses disslosed in mis documeat.
i The informenon contemed herein le for the solo use of Customer ,
4 in oraler to ovoid impearment of riWres of Eamon Nucieer Company, Inc. I
. in poennes or inwoncone which may be ineluded in the informadon contamed in tie document, the fuespeent, by its asempenes of this document ayees not to publish or make public use (in the poesnt use of the arm) of mad informenon until so authonsed in wrvans try Eamon Nuedeer Company, Inc. <
or until after sia (6) nienths following termination or espersaan of the aforesend Ayemment and any extension thereof, unisse otherwise expressly provided in the .", _.._ __ No rights or lienness in or to miy potems are implied by the fumishing of eis desument.
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i XN-NF-86-38 1
l TABLE OF CONTENTS
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s SECIIDN EAGE I,
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1.0 INTRODUCTION
.......................................... I 1
2.0
SUMMARY
............................................... 2 3.0 JET PUMP BWR ECCS EVALUATION M0 DEL.................... 6 3.1 LOCA Description...................................... 6 3.2 EXEM/BWR Appl i cation to Grand Gul f. . . . . . . . . . . . . . . . . . . . 7 i
i 4.0 ANALYSIS RESULTS...................................... 12 4.1 System Analysis....................................... 12 I
4.2 MAPLHGR Results....................................... 13
5.0 REFERENCES
............................................ 45 e
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I 11 XN-NF-86-38 I
LIST OF FIGURES FIGURE PRE 3
2.1 MAPLHGR vs. Average Planar Burnup for ENC 8x8 Reload Fuel in used Grand Gulf Unit 1 LOCA Analysis... 4
~
3.1 System Blowdown Nodal 1zation.......................... 9 3.2 Hot Channel Nodal ization. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 3.3 System Refill /Reflood Nodalization. . . . . . . . . . . . . . . . . . . . 11
! 4.1 Blowdown System Pressure.............................. 12 4.2 Bl owdown Total Break Fl ow. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 l
s 4.3 Blowdown Average Core Inlet Flow...................... 20 4.4 Bl owdown Average Core Outl et Fl ow. . . . . . . . . . . . . . . . . . . . . 21 4.5 Bl owdown HOT CHANNEL Inl et Fl ow. . . . . . . . . . . . . . . . . . . . . . . 22 4.6 Bl owdown HOT CHANNEL Outl et Fl ow. . . . . . . . . . . . . . . . . . . . . . 23 4.7 Blowdown Intact Loop Jet Pump Drive Flow. . . . . . . . . . . . . . 24 4.8 Blowdown Intact Loop Jet Pump Suction Flow. . . . . . . . . . . . 25 i
! 4.9 Blowdown Intact Loop Jet Pump Exit Flow............... 26 4.10 Blowdown Broken Loop Jet Pump Drive Flow. . . . . . . . . . . . . . 27 f 4.11 Blowdown Broken Loop Jet Pump Suction Flow............ 28 4.12 Blowdown Broken Loop Jet Pump Exit Flow............... 29 4.13 Blowdown Upper Downcomer Mixture Level................ 30 4.14 Blowdown Middle Downcomer Mixture Level............... 31 4
i
J iii XN-NF-86-38 LIST OF FIGURES (Continued) l FIGURE EAE 4.15 Blowdown Lower Downcomer Mixture Level................ 32 4.16 Blowdwon Upper Plenum Liquid Mass..................... 33 4.17 Blowdown Upper Downcomer Liquid Mass.................. 34 4.18 Blowdown Lower Downcomer Liquid Mass.................. 35 h
4.19 Bl owdown Lower Pl enum Liquid Mass . . . . . . . . . . . . . . . . . . . . . 36 4.20 Refill /Refl ood System Pressure. . . . . . . . . . . . . . . . . . . . . . . . 37 4.21 Refill /Reflood Lower Plenum Mixture Level . . . . . . . . . . . . . 38 4.22 Refill /Reflood Relative Core Midplane Entrainment..... 39 4.23 Blowdown HOT CHANNEL Heat Transfer Coefficient In MAPLHGR Analysis................................... 40 4.24 Slowdown HOT CHANNEL Center Volume Quality In MAPLHGR Analysis................................... 41 4,J 4.25 Blowdown HOT CHANNEL Center Volume Coolant Temperature In MAPLHGR Analysis................................... 42 l 4.26 Typical Hot Assembly Heatup Results, 80L.............. 43 I
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LIST OF TABLES r
TABLE EASE 2.1 MAPLHGR Results for ENC 8x8 Reload Fuel in Grand Gul f Uni t 1. . . . . . . . . . . . . . . . . . . . . . 5 1
4.1 Grand Gul f Reactor System Data. . . . . . . . . . . . . . . . . . . . . . . . 15 s
4.2 LOCA Event Times for Limiting Break................... 17
- l. 4.3 LOCA BOL Results for Limiting Break................... 18 I
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, 1 XN-NF-86-38 1
1
1.0 INTRODUCTION
i 1
The results of a Loss of Coolant Accident (LOCA) - Emergency Core Cooling System (ECCS) analysis for the Exxon Nuclear Company (ENC) 8x8 fuel in the Grand Gulf Unit I reactor are summarized in this document. The results are presented in terms of the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limit as a function of planar average exposure. These calculations were performed with the generically approved Exxon Nuclear Company EXEM/BWR Evaluation Model(1,2)
[ .
The NRC review of EXEM/BWR(3) which encompassed ,
l application to all jet pump BWR's, concurred with ENC that the model was in -
compliance with Appendix K to 10 CFR 50; hence these analyses comply with Appendix K and satisfy the criteria of 10 CFR 50.46.
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,l A generic BWR/6 LOCA break spectrum analysis that is applicable to Grand Gulf
, has been previously reported (4) . This analysis showed that the limiting break
'l using ENC methodology for the Grand Gulf BWR/6 unit is a guillotine break on the discharge side of the recirculation pump with a break area equal to the
- flow area for both ends of the broken pipe and a discharge coefficient of one.
! For the limiting break defined by the generic BWR/6 break spectrum, the
), LOCA-ECCS analysis described in this report was performed to support reactor
- operation within the expanded power / flow operating map up to 100% of rated thermal power and up to 105% of rated core flow.
i Significant LOCA-ECCS margins were anticipated for Grand Gulf Unit I because i of the unique features incorporated into the BWR/6's: The five nozzle high efficiency jet pumps allow for low pipe break area / vessel volume ratios resulting in longer blowdown. Longer blowdowns allow decay haat rates to significantly decrease before periods of reduced heat transfer are '
encountered. The High Pressure Core Spray System (HPCS), which injects coolant into the reactor vessel upper plenum, provides additional cooling .
compared to the BWR/3 and 4. In addition, the Low Pressure Coolant Injection
- (LPCI) System injects coolant into the bypass region where it is more effective in reflooding the core than in other classes cf BWR plants.
9
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2 XN-NF-86-38 1
2.0 SUPf1ARY l
L For the limiting break, Peak Cladding Temperatures (PCT) were calculated for the reactor flows which t, racket the most severe peaking at rated power: 105 percent of rated core flow and 85 percent of rated core flow. The highest flow permitted in the operating map is 105 percent at 100 percent of rated
, thermal power. At flows below 85 percent, the flow dependant MCPR limit significantly lowers the allowable power of the hot assembly. The operating 3
conditions of 105 percent rated flow resulted in the highest PCT. Therefore, 1
the MAPLHGR analysis was performed at the more limiting operating condition of -
- 102 percent power and 105 percent core flow.
I
' For the limiting break and this limiting operating condition, the exposure dependent MAPLHGR limit was determined for ENC fuel from Beginning of Life (BOL) to a average planar exposure of 45 GWd/MTU. The MAPLHGR for ENC fuel use for this analysis is presented in Table 2.1 and Figure 2.1. The MAPLHGR of Figure 2.1 is constant at 14.3 kw/ft for average planar exposures from 0 to 20 GWd/MTU, and decreases linearly from 14.3 to 9.0 kw/ft for average planar exposures from 20 to 45 GWd/MTU. As anticipated, significant margins (500*F
, in PCT) exist due to the ECCS features incorporated into the BWR/6 Grand Gulf Unit. The limit applies to the initial and subsequent reloads incorporating ENC 8x8 fuel (5) of the design analyzed at reactor powers up to a rated thermal power of 3833 MWt and 105 percent of rated core flow. The MAPLHGR used in this LOCA analysis bounds the more restrictive MAPLHGR limit which is consistent with LHGR limits.
j, All of*these calculations were performed according to Appendix K of 10 CFR 50
([ and the MAPLHGR of Figure 2.1 satisfies the requirements specified by 10 CFR 50.46 of the U.S. Code of Federal Regulations. Operation of the Grand Gulf
,. Unit I reactor with ENC 8x8 fuel within the limit of Figure 2.1 assures that the Emergency Core Cooling System for Grand Gulf Unit I will meet the U.S.
t j NRC acceptance criteria for breaks up to and including the double-endef '
s severance of a reactor coolant pipe. That is: -
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3 XN-NF-86-38 U
- 1) The calculated peak fuel element clad temperature does not exceed the 2200 F limit.
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- 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.
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- 3) The cladding temperature transient is terminated at a time the core geometry is still amenable to cooling. The hot fuel rod cladding oxidation limit of 17% is not exceeded during or after quenching. .
- 4) The system long term cooling capabilities provided for the previous core remains applicable to ENC fuel.
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! Grand Gulf Unit 1 LOCA Analysis
5 XN-NF-86-38 ,
Table 2.1 1
MAPLHGR Results for ENC 8x8 Reload Fuel in Grand Gulf Unit 1
[
Average Planar Local Peak Clad Burnup MAPLHGR MWR Temperature (GWd/MTU) (kw/ft) (cercentJ_, (*F) 1
- 0. 14.3 0.4 1738 5.
10.
14.3 14.3 0.3 0.3 1685 f
1678
- 15. 14.3 0.3 1687 l
- 20. 14.3 0.3 1680
- 25. 13.2 0.3 1642 30.
35.
12.1 11.1 0.2 0.1 1575
[
1496
- 40. 10.0 0.1 1403 (
- 45. 9.0 0.1 1321
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Metal Water Reaction -
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3.0 JET PUMP BWR ECCS EVALUATION MODEL 3.1 LOCA Description 1
f A Loss of Coolant Accident is defined as a hypothetical rupture of the reactor coolant system piping, up to and including the double-ended rupture of the f
q 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 Grand Gulf plant, reactor system coolant inventory loss would result in a
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high containment drywell pressure concurrent with low reactor water level.
These two events would provide a safety signal which would bring emergency h coolant injection systems into operation to limit the accident consequences.
f During the early phase of a LOCA depressurization transient, core cooling is provided by the exiting coolant inventory. Later in the reactor system l depressurization, the High Pressure Core Spray (HPCS) adds to the core heat removal. In the latter stage of system depressurization and after
! depressurization has been achieved, the Lower Pressure Core Spray (LPCS) f provides spray coolant in addition to the HPCS contribution. The Low Pressure Coolant Injection (LPCI) in the bypass supplies liquid to rapidly refill the i bypass. The subcooled liquid being injected into the bypass rapidly fills the l bypass to drive water into the lower plenum through the lower core support structure leakage paths and the bypass holes drilled in the tie plates of the l
fuel. This promotes rapid filling of the lower plenum and early reflooding of the core. During the core reflood process, cooling is provided above the mixture level by entrained r: flood liquid and below the mixture level by pool
. boiling processes.
I The most limiting single failure, assumed in this analysis, was the loss of the diesel generator serving as the power source for two of the three LPCI pumps. This assumption is consistent with the FSAR analysis for Grand Gulf l +
Unit 1. The remaining operational emergency core cooling systems were HPCS, LPCS, automatic depressurization system (ADS) and one LPCI pump.
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i 7 XN-NF-85-38 1
3.2 EXEM/BWR Acolication to Grand Gulf The EXEM/BWR codes were used for the LOCA analysis for Grand Gulf Unit 1. The i
l versions of the EXEM codes used in this MAPLHGR analysis are the same as used in the ENC BWR/6 break spectrum analysasI4) . EXEM/BWR is comprised of the RODEX2, RELAX, FLEX, and HUXY/BULGEX computer codes.
The initial stored energies for the RELAX system blowdown, RELAX / HOT CHANNEL l
and HUXY/BULGEX calculations are determined with the RODEX2 code.
l The RELAX code is used to calculate the reactor system behavior auring the ,.
major portion of the reactor system depressurization transient. RELAX predicts mass distribution, core and system thermal hydraulics, and break flow rates. The blowdown calculation provides core bcundary conditions for the heatup calculation and reactor coolant system conditions for the initialization of the refill /reflood transient calculation using the FLEX coda. The RELAX / HOT CHANNEL heat transfer boundary condition is used in the heatup model up to the time when the LPCS reaches rated flow. The reactor core is modeled with heat generation rate determined from reactor kinetics {
equations with reactivity feedback and decay heating 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 3.2. Pump performance curves characteristic of the recirculation pumps are used in the analysis.
l 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 RELAX / HOT CHANNEL analysis used time dependent plenum boundary conditions from the RELAX system blowdown calculation. The calculated results from the RELAX / HOT CHANNEL calcalation are used as input data to the subsequent hot assembly heatup calculation.
l Conservative heat transfer coefficients and fluid thermodynamic properties for I
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l 8 XN-NF-86-38 I
the heatup calculation are assured by using the maximum stored energy in the i RELAX / HOT CHANNEL calculation for the generation of this information. The o
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
( The FLEX system refill /reflood analysis predicts the latter segment of the reactor coolant system depressurization, the refilling of the reactor vessel 4
lower plenum and the reflooding of the core. The time of hot-node reflood is detemined by FLEX and is an input quantity to the hot assembly HUXY/BULGEX .-
heatup calculation. The FLEX system refill /reflood nodalization is shown in Figura 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 i
spray, decay power, heat transfer coefficients and coolant thermodynamic 7 properties from RELAX / HOT CHANNEL; and time of hot-node reflood and time of i bypass reflood from FLEX. These data are input to HUXY/BULGEX to determine the peak clad temperature and the cladding oxidation percentage. Bounding fission and actinide product decay heat obtained with end-of-cycle neutronics l in the system blowdown calculation assure that the power input to the HUXY/BULGEX heatup calculation is conservative. Conservative heat transfer
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coefficients and fluid thermodynamic properties for the heatup calculation are
- assured by using a bounding stored energy for all exposures of the ENC 8x8 l fuel in the RELAX / HOT CHANNEL calculation.
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11 XN-NF-86-38 1
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. 12 XN-NF-86-38 1
4.0 ANALYSIS RESULTS i
I 4.1 LOCA Analysis An ECCS analysis calculation of the limiting break has been performed for Grand Gulf Unit 1, and the calculated results are summarized herein. 1he break spectrum calculations applicable to Grand Gulf Unit I have been previously performed and reported (4) . The limiting break was found to be a guillotine break on the discharge side of the recirculation pump with a break area equal to the flow area for both ends of the broken recirculation pipe and -
a discharge coefficient of one.
Because of the significant range of allowable flows at full-power, the most limiting full-power flow condition needed to be established. PCTs were calculated for the reactor operating conditions which bracket the most severe peaking condition at rated power: 105 percent of rated flow and 85 percent of
'i rated flow. One hundred five percent of rated flow is the highest flow
- permitted in the operating map at 100 percent power. At core flows below about 85 percent, the reduced flow MCPR thermal limits lower the power (radial peaking) of the limiting assembly. These reactor operating conditions are summarized in Table 4.1. For both operating conditions the hot assembly conditions were the same and chosen at conservatively low values of minimum lg critical power ratio (MCPR) and representative values of the MAPLHGR.
The calculated LOCA event times and results for a limiting break occurring
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when the reactor is operating at the Table 4.1 conditions are summarized in
,I Table 4.2 and 4.3. The PCT is slightly higher for the 105 percent core flow case. Therefore, the 100 percent rated power with 105 percent rated core flow .
is the limiting operating condition for the MAPLHGR analysis, and this power
,j and flow condition is applicable to the extended operating domain. The 105 l percent rated flow case results in a higher PCT because the radial peaking j factor is higher (consistent MCPR assumption) and the vessel depressurizes faster which gives the fuel higher decay power when the refill /reflood heat l1
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l 13 XN-NF-86-38 i
I transfer begins. The 105 percent case depressurizes faster because the recirculation control valve is open farther than for the 85 percent flow case.
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The most significant parameters for the limiting break (1.0 DEG/RD) initiated from the limiting operating condition (105% flow /100% power) are shown as follows: Figures 4.1 through 4.19 are system blowdown parameters, Figures g 4.20 through 4.22 are system refill /reflood parameters. 5 The LOCA system behavior is determined by the system geometry and the break characteristics. Core parameters have only a secondary effect on system event
,. f times. For these reasons, the system analysis descrioad in this report is applicable to future cycles unless system modifications or revised operating conditions negate the plant conditions used herein.
4.2 MAPLHGR Results The MAPLHGR curve analyzed in this report is greater than the MAPLHGR values corresponding to the ENC 8x8 LHGR curve shown in Reference 5. That is, the MAPLHGR curve shown in Figure 2.1 is bounded by the design LHGR limits from
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the fuel rod mechnical design analysis.
The MAPLHGR analysis PCT results for the Grand Gulf Unit I reactor are based on the limiting break initiated from the limiting operating conditions f described in Section 4.1 above. This system analysis used plant specific data applicable to the Grand Gulf Unit 1. MAPLHGR results are obtained using LOCA l
system analysis boundary conditions but require an additional RELAX /H0T
! CHANNEL calculation and a series of RODEX2 and HUXY/BULGEX calculations at {
various fuel exposures. 8 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, and the axial and radial peaking factors l E
F h
1 14 XN-NF-86-38 are consistent with a MCPR of 1.15 and a MAPLHGR of 14.3 kw/ft. This bounding l HOT CHANNEL calculation provides heat transfer coefficient, fluid temperature, l and fluid quality at the plane of interest for the HUXY/BULGEX calculations.
These HOT CHANNEL calculated parameters 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.
i The HUXY/BULGEX calculated results and the corresponding MAPLHGR for ENC 8x8
( reload fuel are shown in Table 2.1 and Figure 2.1. These results conform to I the U.S. NRC requirements specified by 10 CFR 50.46. Table 2.1 shows the e average planar burnup of the hot assembly, MAPLHGR, peak local metal-water reaction and peak cladding temperature. The BOL PCT of 1738'F in Table 2.1 '
for the MAPLHGR analysis is higher than the BOL PCT of 1615'F in Table 4.3
{ because a higher MAPLHGR was used in the final analyses shown in Table 2.1.
It should be noted that because of the ECCS features incorporated into BWR/6's, including Grand Gulf Unit 1, about 500*F of margin to the 2200*F PCT limit is available to account for possible future plant operational changes or other items that might impose a small change in the analysis results.
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i 15 XN-NF-86-38 Table 4.1 I
Grand Gulf Reactor System Data 100% Rated Power / 100% Rated Power /
Event 105% Rated Flow 85% Rated Flow Primary Heat Output, MW 3909.66 3909.66 Total Reactor Flow Rate.
Mlb/hr 118.13 95.63
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l Active Core Flow Rate, 3 Mlb/hr 106.28 85.66 {
Dome Pressure, psia 1049.4 1049.4
.Mid Core Pressure, psia 1066.8 1064.0 Reactor Inlet Enthalpy, g 8tu/lbm 529.2 524.2 g Recirculation Loop Flow Rate, M1b/hr 17.08 14.11
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Steam Flow Rate, Mlb/hr 16.9 16.9 Feedwater Flow Rate, Mlb/hr 16.9 16.9 Rated Recirculation Pump Head, ft 765. 765.
f Rated Recirculation Pump Speed, rpm 1785. 1785.
Moment of Inertia, lbm-ft**/ rad 19,700 19,700 4 I
Recirculation Suction Pipe I.D., in 21.8 21.8 Recirculation Discharge .
Pipe I.D., in 21.8 21.8 I
. E 3833 MW (1.02) = 3909.66 MW h
16 XN-NF-86-38 i
l Table 4.1 I,
(Continued) i Grand Gulf Reactor System Data 100% flated Power / 100% Rated Power /
fY101 105% Rated Flow 85% Rated Flow i
Exposure BOL BOL MCPR 1.15 1.15 '
MAPLHGR 13.2 13.2 -
Axial Peaking Factor 1.39 1.52 Radial Peaking Factor 1.59 1.46' i
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1 17 XN-NF-86-38 Table 4.2 i
LOCA Event Times in Seconds for Limiting Break Reactor Operating Conditions i
100% Rated Power / 100% Rated Power / y Event 105% Rated Flow 85% Rated Flow {
Start Initiate Break 0.00 0.05 0.00
~
l 0.05 Feedwater Flow Stops 3.05 3.05 Steam Flow Stops 6.05 6.05 Low Mixture Level for HPCS 9.54 9.44 Jet Pumps Uncover 17.52 17.40 HPCS Flow Starts 19.54 19.44 Lower Plenum Flash Starts 21.19 20.44 f
Recirculation Suction Uncovers 26.9 27.19 LPCI Flow Starts 91.54 99.10 End of Blowdown (upper plenum y pressure for rated LPCS flow) 110.4 122.0
)
Depressurization Ends (vessel pressure reaches 1 atm.) 170.4 188.0
{
Start of Reflood (high density fluid enters core) 179.4 191.1 l Peak Clad Temperature Reached 190.5 199.0 l
(
18 XN-NF-86-38 I
Table 4.3 c
LOCA BOL Results for Limiting Break i
Reactor Operating Conditions 100% Rated Power / 100% Rated Power /
- Event 105% Rated Flow 85% Rated Flow i Peak Cladding, Temperature l 'F 1615 1598 ,
Peak Temperature Axial Location, ft 6.25 6.25 Local Zr/ Water Reaction (Max.), percent 0.23 0.21 Total Hydrogen Generated, percent of Total Zr Generated <1 <1 i
i I t
. . , - . , . . - , - , - ....,.,....-,-.---.-,-.._.,,,,,-,,_-,.-,c-,n--,-,. - , , - - . - - - - - --.--- - -._-. --... , - -,--- ,.-.
19 7
XN-NF-86-38 i
I
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i
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- i A e a e 3 a P d 20 40 60 80 100 120 140 ISO x
4 TIME (SEC) =.
1 z
m.
Figure 4.2 Blowdown Total Break Flow 8
( L>
I 1
i 1 m 1
0 ?" a g p
. g e
+
p o
o
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. n 4 1
0 a 2 1
0 a 0 1
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i
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i i
l CRAND GULF UNIT 1 i ,
i i i i i i i G - -
1 W+
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i .
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i
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>- N 3 N O
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me O~
u
- -a - - - _ _
i t i e a : a e x
"ib 20 4o so ao too 120 i4o iso ?
l 8' TIME (SEC) in j t a m
i Figure 4.4 Blowdown Average Core Outlet Flow Es i
i a
I i
e I
m
1 23 XN-NF-86-38 e
h l
I
. . . . . 3
. . u
, I
. 8 g
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os (03S/87) et oc or on
'M073 131NI 13NNYH3 10H o oS
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1 24 XN-NF-86-38
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1
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i o~
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if z
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i i
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I I
j g . . . . . . .
, 70 20 40 SO SO 100 120 340 180
- T1NE (SEC) i s.
m m.
i Figure 4.8 Blowdown Intact Loop Jet Pump Suction Flow I
i W
CD l
j '-
i t
I ~
i CRAND CULF UNIT 1 i . . i i i i I
w-i us N
- e dl - -
i $~
o i
.3 IL l
wl x
W m u
n.
J4
~
o i
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l . To 20 40 so so too 120 iso iso TIME (SEC) E
!l i
,L l Figure 4.9 Blowdown intact loop Jet Pump Exit Flow lo I m f 9 g
li1'lll1\ 1lll 1llllll'l1l il\ l i
@ y47mT,m . '
o o
. . . - - ~ t 0
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1
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r D
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l T
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GWmsad 2OJ.a w>_E a.n am L
CRANILGULF UNIT I i
1 i i i i i i GI wg - .
m-N m
d sk o-
.J L&.
=
0 -
W o
s ro m e IL a
m i.o ao e
40 a
so a a so e
too iao a
140 a
soo y '
TIME (SEC) y w
l Figure 4.11 Blowdown Broken Loop Jet Purpp Suction Flow l
l g ' g, e ,
. . .= === - - - - ~* - - -. ... - - .... - w i
1 2
1 gr CRAND GULF UNIT 1 y i I I I I I i "
a
-h u
w-i *
- N
- e
.; Je j g . -
! 2 i o i J
' E j
wIN - -
- x w W o i L i 7 . -
! Jo l m -: --
1 i
~ ~
l l
i e a a e a a e s x
, 'ib to 40 so so 100 120 140 Iso ?
TIME (SEC) %.
m m
i L
m
,i Figure 4.12 Blowdown Broken loop Jet Pump Exit Flow u
I l
i GRAND GULF UNIT 1
,, i . . . i i i N
1 gg ta.
I i J l w.
w 4
i
.J
! X
) E N
~
u Q
O a:
w i
- a. . .
m.
o i
j t I a a e a a a M
ca 20 40 so so too 120 140 iso [
TINE (SEC) T
,1 8, w
m Figure 4.13 Blowdown Upper Doncomer Mixture level j
i M SP4 WW F45 M IEWS #4B Wuut muW EuS Muut mim em Wee Wee Sus suum amme e
i
)
i i
GRAND GULF UNIT 1 e.
i i i i i i .
o .
n-e-.
I 14.
i w
\
.,a -
we w
l J
i e x 1 - -
i E.
o u o m l
O
! E,
! .t i -
, w 1
1 1
I ?
z 4 a e i e
' i n
,n i es to 40 so so 100 i20 140 iso g TlHE (SEC) j a
m 1
i i
l l Figure 4.14 Blowdown Middle Downcomer Mixture Level il
i l
i GRAND GULF UNIT 1
,, i i i i i i i N
~ "
C2 i
is.
w ll J w,
i >. 1 i 14 l
- x l r, _ _
o~
O w w
a:
w 2 _ _
, O, a
i _ _
i x a a e a a e s 2 cc ao 4o so so too iao i4o iso =
1 '
- T1HE (SEC) a, o
b Figure 4.15 Blowdown Lower Downcomer Mixture level W M
g_
. w* xP" imawe n n
_ o
_ _ ~ - -
s i
0 n 4 1
o
. a t s
0
. a 0 1
) s s
C a E M S d
- ( i
- . a o uq sE i M L I
T m z l u
n e
. i o P 1 s
. r
-. e T
I p
. p N U U n
. n 0 w
- F 4 o L d U w o
G l
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.- D
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A t R 4
- G
- e r
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_ _ ' . - F ec t
I.
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. .mm<r o 3O .
J. E3zwJA ,wAL3 11l4Iii!; < - i-* ,i!'4111111 1;4 Jll : ' !ij. !Ii : 9i
35 XN-NF-86-38 .1 I
I i i i . .
2 I
I E
l I
I
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n x 0 m w -
S E
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r 5-u 5 E g 8
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- =
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3 i w -
J 8 o s a o E
I a
. ~
i . .
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e f
s
? ,
- ,e .- - .i - si -. -,
(81) 'SSVW 010011 H3WOONM00 H3ddn 5 I
t
& E,5da a, 1l o
s
- - - - - i 0
3 , 4 1
0 3 . 2 -
1 o
y
. o t
)
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S s
( s
. I . o a sE M M d i
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i L
y . o r f 1 T
s e m
o I c n
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. 0 4 r L e U w G o L
D 8 N g . 0 1
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u g
i F
r - - - - -
n 9ga og. . ec
- s ams- .
u Er O mO ' cwrouz20O c mw3Os_
i1 1 ! Ii,;I!1i: !;:
,i > i .. ' :!;i4i!')11 !,! ,il, !:(l ii <
I L-37 XN-NF-86-38 I
I
-S I
! I
- R .-
I g
G.3
$ .Ia 1
- S w g i
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s-- O o.
0
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i
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o i C
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-g 3
x
=
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(W I I -
M t i i
00006 00000 00004 1 0000EL 0000l1 00000L (91) *SSVW 010011 WAN31d W3M01 i
E I
l f
m
_ .. . __ .- .6.. .
_. _ __i i
i i
GRAND GULF UNIT 1 i i i i i i i l g
_g . -
U-5
.n i
f
".E g a -
\ w m
1 a 1 en
! m . -
i w
- e i n. w
- co i 5+
a i Mg i - -
l a
i 1
1 4
1 x
n a a a a a n 2
j i
e 110
~
130 150 170 190 210 230 250 270 k TIME FROM BREAK INITIATION (SECS.) h 1
6 i
- Figure 4.20 Refill /Reflood System Pressure i
t I
i GRAND GULF UNIT 1 l ,
i i i . . . .
ll u,
f t~
x 1 .
I. _.
w -
C, .
X i -
4 E -
! En
! 3 z to w e J
. 1 -
I m*
w 2
o 1 -I l -
i n
DC I O I I R R R T
2
, n 110 130 150 170 190 210 230 250 270 [
TIME FROM BREAK INITIATION (SECS.) _. . .
T a
Figure 4.21 Refill /Reflood Lower Plenum Mixture Level
_ .O y!=TRb
_ 0 1
- - - - . 2 I 6 0
2 t
n e
2 m i a 0 i n .
2 a r
t n
E e
8 n i i 9 a
- l 1 p
_ )
. d i
S M C e
. E r
- 4S 9( o
_ i a 1 C E e
_. v H i t
1 T
a l
- _ e i I 0 R 9
1
. d o
. ._ 1 l o
f
- T I e
- N R 6 /
U i i 8 l 1 l F i L f e
U R G
2 2 D B 8 2 N i 1
A 4 R e G r u
g 8 i
- 7 F
, - - - - i 1
- 5. .
- - *o *i ,4 9o g&
~ wzwE zw wz<a_a.a r w> "<Jwa
I I
i GRAND cut.F Unit 1 .
j %-
E i i i i i , , .
i w :
3 o I O 1 z O j
- 1 .
- z -] .
w .
- o .
1 e I A
>j h
o-5 o : :
! 4 i A n
vs . V .
z a
< ~
i E% i:
s h 3
! w ~
I I t l
i
! %- 5 :
I : -
S t a a t a i a f . b 20 40 so so too 120 140 iso x i
TIME (SEC) T
=,
Figure 4.23 Blowdown HOT CHANNEL Heat Transfer Coefficient in MAPLHGR Analysis y' 03 l
j N NE G CG .7., .%- y-~
~
s==m-.
= . amma seum og g a g g 4,
i i
1 i
i
! CRAND CULF UNIT 1 Sr i i i i . . .
i
- o . .
>E e-w
]
.i 8ea - -
i w q O
1 0 z
ec . - -
wd l
E A w ro l o 1l oo - .
j Id 1
l l
M - ~
O l
1 4
,t a e a a a a e x
! , & 2o so so ao too 12o 14o iso ?
l TlHE (SEC) 4
! =
! Figure 4.24 Blowdown 'l0T CHANNEL Center Volume Quality in MAPLHGR Analysis
43 XN-NF-86-38 I
I O
I
, , , i ' ~
g S
%u g l G) c.
n 5 ..
c n o v c
O I
- n 8o O "o W >
us L w
- a 3c
=w r
3
~ d E E 5 g 22, v g
^
- 8a4c z e D s a C
S 88 ,
b 8i
- 5t a
z -
g-n.
x O
2 e
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{
t oos oo, od. 3 oos nos ooz ooo g (3 030) 'dW31 INY1000 10A 831N30 I.
I Ii I
,a ;
~
j . ._ .T_. ._ . . . . . ._ .. _ __ _ .. -
! GRAND GULF UNIT 1 l
l i
i 1 2345678 I S 101112131415 l 3 10161718192021 j 4 1117 2223242! 26 l 5 121823272921'30 6 13IS2428313233 7 14202529323435 8 15212630333536 2500-
- PIN 4
, O PIN 11 O CANISTER i C e 2000 -
g
! E t
m i $
l g 1500 -
i
- l
- W -
I o
1000 - m o
]
. a- t
-" x .
l 500F A { ,
U- b a t i t t 9 I I a
O 40 80 120 160 200 240 280 320 Tit 1E (SEC) -
Figure 4.26 Typical Hot Assembly Results, BOL ,
. j
t -
45 XN-NF-86-38 I
5.0 REFERENCES
I
- 1. " Generic Jet Pump BWR 3 LOCA Analysis Using the ENC EXEM Evaluation Model", XN-NF-81-71(A), Supplement 1, Exxon Nuclear Company, September 1982.
{*
- 2. " Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model",
XN-NF-82-07(A), Revision 1, Exxon Nuclear Company, November 1982.
- 3. " Exxon Nuclear Methodology for Boiling Water Reactors: EXEM ECCS Evaluation Model", XN-NF-80-19fA), Volume 2, Revision 1, Exxon Nuclear Company, June 1981.
- 4. " Generic LOCA Break Spectrum Analysis For BWR/6 Plants", XN-NF-86-37(P),
Exxon Nuclear Company, April 1986.
- 5. " Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel",
XN-NF-85-67(P), Revision 1, Exxon Nuclear Company, March 1986.
- 6. " Grand Gulf Unit 1 Cycle 2 Reload Analysis", XN-NF-86-35, Exxon Nuclear Company, May 1986.
t l
l 1
I l
I i
f
l l
XN-NF-86-38 Issue Date: 5/30/86 GRAND GULF UNIT 1 LOCA ANALYSIS Distribution DJ Braun JC Chandler RE Collingham RA Decker SF Gaines TL George (NAI) <
- g CE Hendrix (ITI)
JG Ingham SE Jensen JE Krajicek TL Krysinski JN Morgan '
DA Prelewicz (ENSA)
GL Ritter DR Swope HE Williamson l MP & L/JD Floyd (40) t Document Control (5) i 1
e i
. _ _ _ , . _ . . - . . _ . , _ _ _ _ , . . _ , . . . _ _ . . . . _ _ _ , _ _ _ , _ . _ _ _ _ _ _ _ . . _ _ . _ _ _ _ , _ . _ . _ _ . _ _ _ _ _ , _ . _ _ _ , _ . . _ _ _ _ , , _ _ _ . _ _ _ _ _ , . . - . . _ , . _ . _ . _ _ _ _ _ _ _ . _ _