ML17146A422

LOCA-ECCS Analysis MAPLHGR Results for 9x9 Fuel.

ML17146A422
Person / Time
Site:	Susquehanna
Issue date:	05/31/1986
From:	Collingham R, Swope D, Ward G SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17146A415	List: ML17146A414 ML17146A416 ML17146A417 ML17146A419 ML17146A421 ML17146A422 ML17146A423 ML18040B150
References
XN-NF-86-65, NUDOCS 8606240325
Download: ML17146A422 (58)

Text

XN-NF-86-65 SUSQUEHANNA LOCA-ECCS ANALYSIS MAPLHGR RESULTS FOR 9X9 FUEL MAY 1986 RICHLAND, WA 99352 EXXON NUCLEAR COMPANY, INC.

8b0b240325 Sb0b19 PDR ADOCK 05000388 P PDR

l XN-NF-86-65 Issue Date:

5/9/86 SUSQUEHANNA LOCA-ECCS ANALYSIS NAPLHGR RESULTS FOR 9x9 FUEL Prepared by: O'W Z5

~ ~ Mope BXR Safety Analysis Concur:

o ng , anager BMR Safety An ysis Concur:

ar , anager Reload Licensing Concur:

organ, ager C tomer Services Engineering .

Approve: wAPg samson, anager Licensing 4 Safety Engineering Approve:

er, anager Fuel Engineering L Technical Services tmrc

.EQON NUCLEAR COMPANY, INC.

NUCLEAR REGULATORY COMMISSION DISCLAIMER IMPORTANT NOTICE RECARDINQ CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY This technical report was rlerived through research and development programs sponsored by Exxon Nuclear Company, Inc. It is being sub.

mitted by Exxon Nuclea'o the USNRC as pat of a technical contri-budon to facilitate safety analyses by licensees of the USNRC which udlite Exxon Nudear fabricated reload fuel or other technical services provided by Exxon Nuclear for Iiaht water power reactors and it is true and conact to the best of Exxon Nuclear's knowledge, informadon, and belief. The information contained herein may be used by the USNRC in its review of this report, and by licemees or applicants before the USNRC which are customers of Exxon Nuclear in their demonstradon of compliance with the USNRC's regulations.

Vflthout demgadng from the foregoing, neither Exxon Nuclear nor any pe+on acting nn its behalf:

A. Makes any weranty, express or implied. with respect to the accuracy, completeness, or usefulness of the infor madon contained in this document, or that the use of any information, apparatus, method, or process disclosed in this document will not infringe pnvately owned nghts; or B. Assumes any liabiiities with respect to the use of, or for darages resuldng from the use of, any informadon, ap-paratus, method, or process disclosed in this document.

XN. NF- FOO, 766

XN-NFo86-65 TAB 0 CONT NTS

~SECT ON PAG

1.0 INTRODUCTION

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1 2.0 S UMMARY~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 2 3.0 JET PUMP BMR ECCS EVALUATION HODEL..................... 4 3.1 LOCA Description....................................... 4 3.2 EXEH/BWR Application to Susquehanna.................... 4 4.0 ANALYSIS RESULTS.........................:............. 12 4.1 S ystem Analysis........... ............................ 12 4.2 HAPLHGR Results........................................ 14

5.0 REFERENCES

............................................. 43

XN-NF-86-65 IST OF TAB S TABLE PAG 3.1 Susquehanna Reactor System Data........................ 8 4.1 Susquehanna Limiting Break Event Times................. 15 4.2 Susquehanna MAPLHGR Results for ENC 9x9 R eload Fuel............................................ 16

XN-NF-86-65 ST OF FIGUR S FIGUR PAG 2.1 Susquehanna NAPLHGR Results for ENC 9x9 Reload

'I fuel............................... 3 3.1 System Slowdown Nodalization.......................... 9 3.2 Hot Channel Nodalization.............................. 10 3.3 System Refill/Reflood Nodalization.................... 11 4.1 Blowdown System Pressure....... 17 4.2 Blowdown Total Break Flow............................. 18 4.3 Blowdown Average Core Inlet Flow........... ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ 19 4.4 Blowdown Average Core Outlet Flow.......... ~ ~ ~ i ~ ~ ~ ~ ~ ~ ~ 20 4.5 Blowdown Hot Channel Inlet Flow............ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 21 4.6 Blowdown Hot Channel Outlet Flow,.......... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1 22 4.7 Blowdown Intact Loop Jet Pump Drive Flow... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 23 4.8 Blowdown Intact Loop Jet Pump Suction Flow. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 24 4.9 Blowdown Intact Loop Jet Pump Exit Flow.... ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ 25 4.10 Blowdown Broken Loop Jet Pump Drive flow... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 26 4.11 Blowdown Broken Loop Jet Pump Suction Flow. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 27 4.12 Blowdown Broken Loop Jet Pump Exit Flow.... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 28 4.13 Blowdown Upper Downcomer Nixture Level..... i ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 29 4.14 Blowdown Hiddle Downcomer Hixture Level.... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 30

XN-NF-86-65 IST0 F GUR S (Cont.)

FIGUR PAG 4.15 Blowdown Lower Downcomer Mixture Level................ 31 4.16 Blowdown Lower Downcomer Liquid Hass....... ~ ~ ~ ~ ~ 1 ~ ~ ~ ~ ~ 32 4.17 Blowdown Upper Plenum Liquid Pass.......... ~ t ~ ~ ~ ~ ~ ~ ~ ~ 33 4.18 Blowdown Upper Downcomer Liquid Hass....... ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ 34 4 ..19 Blowdown Lower Plenum Liquid Hass.......... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 35 4.20 Refill/Reflood System Pressure............. ~ ~ ~ ~ ~ ~ ~ o 36 4.21 Refill/Re'flood Lower Plenum Hixture Level.. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 37 4;22 Refill/Reflood Relative Core Hidplane Entrainment....................... ~ ~ ~ ~ ~ ~

~ ~ ~ ~ 38 4.23 Blowdown Hot Channel Heat Transfer Coefficient................................. 39 4.24 Blowdown Hot Channel Center Volume guality................................ ~ ~ ~ ~ ~ ~ ~ ~ 40

4. 25 Blowdown Hot Channel Center Volume Coolant Temperature................. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 41 4.26 Typical Hot Assembly Heatup Results, BOL, MAPLHGR 10.2 kw/ft....................... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 42

XN-NF-86-65 The results of a LOCA-ECCS analysis for ENC 9x9 fuel in the Susquehanna reactors are summarized in this document. The results of this analysis are presented in terms of the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limit as a function of fuel exposure. These calculations were performed with the generically approved Exxon Nuclear Company EXEM/BWR Evaluation Model 'ccording to Appendix K of 10 CFR 50',

(3) and the results comply with the U.S. NRC 10 CFR 50.46 criteria.

A break spectrum analysis applicable to Susquehanna Units has been previously accepted by the U.S. NRC for Susquehanna Unit 1 ~ This analysis showed that the limiting break is a double-ended guillotine (DEG) break in the recirculation discharge (RD) piping with a discharge coefficient of 0.4.

A MAPLHGR analysis for ENC 8x8 fuel in Susquehanna Unit 1 was previously reported . These results are applicable to Susquehanna Units 1 5 2 because there are no significant differences between these BWR 4 sister plants and because conservative core conditions were assumed in the analysis. Similarly, the MAPLHGR results being reported for ENC 9x9 fuel are applicable to Susquehanna Units 1 5 2.

XN-NF-86-65 2.0

SUMMARY

A limiting break calculation was performed for Susquehanna plant specific geometry and conditions to establish boundary conditions for the heatup analyses. The exposure dependent MAPLHGR limit was determined for ENC 9x9 fuel from beginning-of-life (BOL) to an assembly exposure of 40 GWd/HTU using.

the limiting break boundary conditions. The resulting HAPLHGR limit for ENC fuel is presented in Figure 2.1. The HAPLHGR limit of Figure 2.1 is constant at 10.2 kw/ft for assembly exposures from 0 to 20 GWd/HTU, decreases to 9.6 kw/ft at 25 GWd/HTU, and then declines linearly to 7.5 kw/ft as the assembly exposure increases to 40, GWd/MTU. The limit applies to the initial and subsequent reloads incorporating ENC 9x9 fuel of the design analyzed. The reported results apply to both Susquehanna units and are bounding for reactor operating conditiops up to 100% rated power and 108% rated flow.

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 Susquehanna reactors with ENC 9x9 fuel within the limit of Figure 2. 1 assures that the Emergency Core Cooling System will meet the U.S. NRC acceptance criteria for 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. The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1/ of the total amount of zircaloy in the reactor.

3. The cladding temperature transient is terminated at a time the core geometry is still amenable to cooling. The hot fuel rod cladding 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.

12 20,10.2)

(25,9.6) 0 10 15 20 25 30 Bundle Average Burnup GMd/HTU Figure 2. 1 Susquehanna HAPLIIGR Results for ENC 9x9 Reload Fuel

XN-NF-86-65 3.0 JET PUMP BWR ECCS V UATION MOD 3.1 LOCA Descri tion A loss-of-coolant accident (LOCA) is defined as a hypothetical rupture of the reactor coolant system piping, up to and including the double-ended rupture of the largest pipe in the reactor coolant system or of any line connected to that system up to the first closed valve. In the unlikely event a LOCA occurs in one of the Susquehanna plants, the reactor system coolant inventory loss would result in a high containment drywell pressure and reduced reactor vessel pressure. The concurrent high drywell pressure and low reactor vessel pressure provide a safety injection signal which brings coolant injection systems into operation to limit the accident consequences.

During the early phase of the LOCA depressurization transient, core cooling is provided by the exiting coolant inventory. In the latter stage of system depressurization and after depressurization has been achieved, the core spray provides core cooling and supplies liquid to refill the lower portion of the reactor vessel and reflood the core. The reflood process provides sufficient heat removal to terminate the core temperature transient.

3.2 BW 1 c tio to S ue a The ENC EXEM/BWR codes were used for this LOCA ECCS analysis. EXEM/BWR is comprised of the RODEX2, RELAX, FLEX, and HUXY/BULGEX computer codes. The following code versions were used in this analysis:

XN-NF-86-65 Code Version RODEX2 UAPR85 RELAX UJUN85 FLEX AAPR86

'HUXY/BULGEX UOCT85 The versions of RODEX2, RELAX and HUXY/BULGEX are the same as were reported in Reference 6. The only change to the discussion of these code versions from that in Reference 6 is the use of the new creep coefficient for Zircaloy-2 Bl<R cladding reported in XN-NF-82-06(P), Rev. 1, Supp. 1 (April 1984). The AAPR86 version of. FLEX is the version used in previous ENC LOCA analyses for Susquehanna units (References 4, 5, and 7), with updates to evaluate 9x9 array fuel. These updates, required to perform the radiation heat transfer calculation for 9x9 fuel assemblies, were previously used and reported in Reference 8.

The initial stored energies for the RELAX blowdown, RELAX hot channel and HUXY/BULGEX calculations were determined with the RODEX2. code.

The RELAX code was used to calculate the reactor system behavior during. the initial 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 boundary conditions for the heatup calculation and reactor coolant system conditions for the initialization of the refill/reflood transient calculation using the FLEX code. Because the FLEX code does not model closure of the valves in the recirculation piping, the system blowdown code, RELAX, was run until both the low pressure core spray reached rated flow and recirculation line valves were completely closed. This approach was necessary to properly model event times and was used in previous ENC Susquehanna analyses; the RELAX hot channel heat transfer was used in the heatup model only until the time that rated core spray flow was calculated. The reactor core was modeled with heat generation

XN-NF-86-65 rates 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 was nodalized into control volumes representing reasonable homogeneous regions interconnected by junctions as shown in Figure 3. 1. Reactor system data for the Susquehanna LOCA system analysis are summarized in Table 3. 1. This data applies to Susquehanna Unit 1 and Unit 2 because of the similarities of these sister plants. Pump performance curves characteristic of the Susquehanna recirculation pumps are used in the analysis.

For the maximum power fuel assembly, a separate RELAX hot channel calculation was used to calculate the cladding-to-coolant heat transfer coefficients and the coolant thermodynamic properties. The hot channel analysis used time dependent plenum boundary conditions from the RELAX blowdown calculation. The calculated results from the hot channel calculation were used as input data to the subsequent hot assembly heatup calculation until the time of rated core spray flow was calculated. Conservative heat transfer coefficients and fluid thermodynamic properties for the heatup calculation were assured by using the maximum stored energy in the hot channel calculation for the generation of this information. The HUXY/BULGEX hot assembly heatup calculation computes the fuel temperature transient from its initiation through peak clad temperature (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 lower plenum and the reflooding of the core. The time of hot-node-reflood was determined by FLEX and is an input quantity to the hot assembly HUXY/BULGEX heatup calculation. The FLEX system refill/reflood nodalization is shown in Figure 3.3.

The HUXY/BULGEX heatup calculation computes the entire LOCA temperature transient and uses: fuel stored energy, thermal gap conductivity and

XN-NF-86-65 dimensions from RODEX2 as a function of power and exposure; time of rated spray, decay power, heat transfer coefficients and coolant thermodynamic properties from RELAX; and time of hot-node-reflood from FLEX. For the spray cooling period the Appendix K spray heat transfer coefficients were used: For heated fuel rods the values of 1.5, 3.0 and 3.5 BTU/hr-ft2 -F were used in the interior, corner and peripheral fuel rods respectively; for the unheated surfaces (channel and water rods) the spray heat transfer coefficient of 5 BTU/hr-ft2 -F was used. For the reflood period the Appendix K reflood heat transfer coefficient value of 25 BTU/hr-ft -F was used. These data were input to HUXY/BULGEX to determine the peak clad temperature (PCT) and the cladding oxidation percentage.

XN-NF-86-65 Table 3.1 Susquehanna Reactor System Data Primary Heat Output, HWt (102% of rated) 3359 Total Reactor Flow Rate, Hlb/hr (108% of rated) 108 Active Core Flow Rate, Hlb/hr 97.4 Reactor Dome Pressure, psia 1033 Reactor Inlet Enthalpy, Btu/lb 524 Total Recirculation Loop Flow Rate, Hlb/hr 31.95 Steam Flow Rate, Hlb/hr 13.78 Feedwater Flow Rate, Hlb/hr 13.75 Rated Recirculation Pump Head, ft 710 Rated Recirculation Pump Speed, rpm 1670 Homent of Inertia, ibm-ft2 /rad 15710 Recirculation Suction Pipe I.D., in 25.34 Recirculation Discharge Pipe I.D., in 25.34

XN-NF-86-66 ADS STSAN'OS 22 Qe Stean Dome 38 Fecdwatar HPCl Qa Upper Downcomer Spray {LPCSl

,7s 71

'7 Qa Upper P4raan tas Q7 171 271 1d 23 0 Containment co ~

CL l

E Q29 la ~I v

ca cs Qll Ql Q17 2c 121 la Lower Plenum lUpperl 19 10 Qa LPCI 29 30 1 22 Ol Guide Tubes C19 tt Lower Plenum Qe Ct lLowcr) ta 15 Q Volun>> Q29 Q21 Qtc Junction Pump Rccircuhdon Rccirculadon Pump isolation Isohtion Valve Valve Figure 3.1 System Blowdown Nodalization

10 XN-NF-86-65 UPPER PLENUM 09 . 8 08 07 06 05 04 03 LOWER PLENUM 0 vOLUME JUNCTION HOT CHANNEL NODALIZATION Figure 3.2

XN-NF-86-65 Steam Dome 4 Upper Downcomer 0>

Upper Plenum L pcs 4

CI Recfrcufatfon C

c-o4 CL Dfacharge Pfpfng B 8

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~ Junctfon SYSTEM REFILL/REFLOOD NODALIZATION Figure 3.3 Xt&CH4II9

12 XN-NF-86-65 A limiting LOCA-ECCS analysis has been performed for the Susquehanna units with all ENC 9x9 fuel and the results are summarized herein. Break spectrum calculations applicable to the Susquehanna Units have been performed and reported in Reference 4. In this analysis the limiting break location was found to be in the recirculation discharge piping and the limiting size and configuration were found .to be a double-ended guillotine (DEG) with a discharge coefficient of 0.4.

The LOCA system behavior is determined by the system geometry, the initial operating conditions, the ECCS and the break characteristics. Core parameters have only a secondary effect on system event times. The ENC 9x9 fuel in this HAPLHGR analysis has 79 fuel rods and 2 inert water-filled rods in each assembly. One of the water rods functions as a spacer capture rod. Seven spacers maintain fuel rod spacing. The fuel rods are pressurized to 5 atmospheres and have 30 mil thick Zircaloy-2 cladding. Each rod contains either UO or UO -Gd 03 at a nominal density of 94.5%. Since the ENC 9x9 fuel analyzed for the Susquehanna units is hydraulically compatible with the ENC 8x8 fuel analyzed in the spectrum analysis, the limiting break identified by the spectrum analysis is applicable to this HAPLHGR analysis.

In addition to analyzing the limiting break, the limiting single failure for the ECCS was modeled and the, LOCA was assumed to occur while the plant was operating at the limiting steady state operating condition. The limiting sing'le failure to the ECCS was identified by the Nuclear Steam Supply System vendor as the failure of the LPCI injection valve to open for the intact recirculation loop. The limiting operating condition was identified as the highest power and the highest flow permitted by the operating map (557) . The limiting operating condition is therefore 1005 rated power and 108/. rated flow. This is limiting because the power distribution in the hot bundle is

13 XN-NF-86-65 based on the HCPR and HAPLHGR limits. To obtain the same HCPR and MAPLHGR in the hot bundle a higher bundle power is required at 108% flow. The reported results are bounding for reactor operating conditions up to 100% rated power and 108% rated flow.

The results of this HAPLHGR analysis are applicable to 9x9 fuel in both Susquehanna Units since there are no significant differences between these sister plants. Since conservative core conditions wer'e evaluated in this analysis, the results are applicable to future cycles of the Susquehanna units unless system modifications or revised operating conditions negate the plant conditions used in the analysis.

Major event times and results for the limiting break for Susquehanna Unit 2 are shown in Tables 4.1 and 4.2. Figures 4. 1 through 4. 19 are system bl.owdown parameters. Figures 4.20 through 4.22 are system refill/reflood parameters.

14 XN-NF-86-65 4.2 MAPLHGR Results The MAPLHGR results for ENC 9x9 fuel in the Susquehanna Units are based on the results of the limiting system analysis described in Section 4. 1. This system analysis used plant data applicable to both Susquehanna units. The MAPLHGR results were obtained using LOCA system analysis boundary conditions, but required an additional RELAX hot channel calculation and a series of HUXY/BULGEX calculations at various fuel exposures.

A bounding hot channel calculation was performed for this MAPLHGR analysis in which the fuel stored energy was the maximum for the exposure range of interest for ENC 9x9 fuel. This bounding hot channel calculation provided heat transfer coefficients, fluid temperature 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.

Target values for MAPLHGR limits were selected by combining the LHGR design limits from the fuel rod mechanical design analysis and appropriate power peaking factors. The analyses performed with these target MAPLHGR limits meet the requirements of 10 CFR 50.46 for exposures above 20 GWd/MTU. For exposures below 20 GWd/MTU lower MAPLHGR limits were necessary to meet the requirements of 10 CFR 50.46. The HUXY/BULGEX calculated results and the corresponding MAPLHGR limits for ENC 9x9 reload fuel are shown in Table 4.2 and Figure 2.1. For exposures below 20 GWd/MTU the MAPLHGR limits are more restrictive than the LHGR design limits. For exposures above 20 GWd/MTU the MAPLHGR limits are consistent with the LHGR design limits. In Table 4.2 exposure is expressed in terms of the average burnup of the hot assembly.

15 XN-NF-86-65 Table 4.1 Susquehanna Limiting Break Event Times Event Time sec Start 0.00 Initiate Break 0.05 Feedwater Flow Stops 0.55 Steam Flow Stops 5.05 Low Mixture Level (LI) 12.9 Jet Pumps Uncover 16.3 Recirculation Suction Uncovers 25.8 Lower Plenum Flashes (guality > 0.) 19.9 HPCI Flow Starts 33.4 LPCS Flow Injection Starts 80.6 Rated Spray Calculated 127.1 Depressurization Ends (Vessel pressure reaches 1 atmosphere) 259.6 Start of Reflood (high density fluid enters core) 273.7 Peak Clad Temperature Reached 286.5

16 XN-NF-86-65 Table 4.2 Susquehanna MAPLHGR Results for ENC 9x9 Reload Fuel Assembly Average Local~* Peak Clad Burnup MAPLHGR MWR Temperature

~

GWd MTU ~kw ~~t ~1o F 0 10.2 3.82 2060 5 10.2 3.63 2069 10 10.2 3.68 2121

'15 10.2 4.76 2140 20 10.2 5.14 2147 25 9.6 2.65 2016 30 8.9 0.99 1839 35 8.2 0.69 1752 40 7.5 0.47 1676 Metal Water Reaction.

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>C oo I ll 100 I 20 140 160 I 10 20 40 60 $0 CO Cll T I NE (SEC) I Ch Vl Figure 4.11 Blowdown Broken Loop Jet Pump Suction Flow

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29 XN-NF-86-65 O

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Figure 4.15 81owdown Lower Downcomer Mixture Level

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hl O

'/0 2/." ?80 2/5 290 295 300 TIME (SECS. )

f-'iqnre 4.22 Refill/Reflond Relative Core'Hi<lplane Fntrainment

SusauEHANNA 0.4 DEG/RD HC

>C I

II 20 40 60 80 100 120 1io 180 I CO TIME (SEC) CJl I

Ch CJl Figure 4.23 Dlowdown Hot Channel Heat Transfer Coefficient

SUSOUEHANNA 0.4 DEG/RD HC 4

I a CO IU ~

K C)

UJ I

UJ C3

~ci EV 0

O 20 40 60 80 IQO )20 140 160 T I NE (SEC)

Figure 4.24 Blowdown Hot Channel Center Volume equality

SUSQUEHANNA 0 4 OEG/RD HC 0

x-oa Lal 1

Ev I

td K

-jo Oo

~ CO 0

ld 0

O+o 4J o

CI OC I

I 20 io 80 80 100 120 110 160 CO CJl TINE (SEC) I Ol CJ1 Figure 4.?5 81owdown Hot Channel Center Volume Coolant Temperature

SUS02 HUXY ICF 0.40EG/RD 2 3 4 5 6 7 8 9 2 lOll 1213 I4151617 I I 18 19 20 21 22 23 24 4 12192$ Z6 2? 282930 5 1320263132333435 6 1421273236373839 7 1522283337404142 8 1623293438414344 2500 9 1724303539424445 PIN26 SSgPIIII

'CANISTER 2000 D

I o 1500 K

IAj 1000 CJ 50 40 80 120 160 200 240 280 320 T l HE (SEC) f fy>re 4.76 Typical Ifnt Rssemhly IfeatiifI ftesul f.s, flOI., MRf'f.ffGR = 10.2 kw/ft

t R

XN-NF-86-65 5.0 1.

2.

REFERENCES "Generic Jet d I",

September

~E--->>,Rid "Exxon

~,

1982.

Nuclear Pump BWR Company 3 LOCA pl ECCS E,EN Analysis Using the I

I, Cladding E

Swelling py,l N

ENC, EXEM and Rupture

.,II p Evaluation Model",

3. "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors", 10 CFR 50.46 and Appendix K of 10 CFR 50,

4. "Generic P

Company, LOCA C I Inc.,

I 'i Federal Register, Volume 39, Number 3, January 4, 1974.

Break Spectrum Analysis December 1984.

Ltd",

for

~F- BWR I,

3 and 4 with Modified E

Low

5. "Susquehanna Unit 1 LOCA-ECCS Analysis HAPLHGR Results", XN-NF-84-119, Exxon Nuclear Company, Inc., December 1984.

N. "LXA B k Np A ly1 Nuclear Company, Inc., December 1985.

1 PER N", ~k- ->>, E

7. "ECCS Analysis For Susquehanna Unit 1 at Full Power- and 86% Flow Using the ENC EXEH BWR Evaluation Model", XN-NF-85-14, Exxon Nuclear Company, Inc.,- February 1985.

8. "Dresden Unit 3 LOCA-ECCS Analysis MAPLHGR Results for 9x9 Fuel",

XN-NF-85-63, Exxon Nuclear Company, Inc., September 1985.

9.

~E"-N--,R.!,EN "Generic Mechanical I Cp,l Design For Exxon Nuclear Jet

.,ApllltReload Pump BWR Fuel",

E I

I

XN-NF-86-65 Issue Date:

5/9/86 SUSQUEHANNA LOCA-ECCS ANALYSIS MAPLHGR RESULTS FOR 9x9 FUEL Distribution D.J. Braun R.E. Collingham S.F. Gaines J.G. Ingham S.E. Jensen T.H. Keheley J.E. Krajicek J.N. Morgan G.L. Ritter D.R. Swope H.E. Williamson T.L. George (Numerical Application)

PPEL/H.G. Shaw (40)

Document Control (5)

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