ML19296D096
| ML19296D096 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 04/09/1979 |
| From: | Cooke G, Jensen S, Valentine P SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML19296D077 | List: |
| References | |
| XN-NF-79-021, XN-NF-79-021-R01, XN-NF-79-21-R1, NUDOCS 8002290334 | |
| Download: ML19296D096 (20) | |
Text
.
g XH NR9 21 REV.1 BIG ROCK POINT LOCA ANALYSIS USING THE EXXON NUCLEAR COMPANY WREM NJP-BWR ECCS EVALUATION MODEL MAPLHGR ANALYSIS APRIL 1979
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XN-NF-79-21 Revision 1 Issue Date:
04/09/79 BIG ROCK POINT LOCA ANALYSIS USING THE EXXON NUCLEAR COMPANY WREM NJP-BWR ECCS EVALUATION MODEL MAPLHGR ANALYSIS Prepared by P. J. Valentine S. E. Jensen G. C. Cooke
/1 4-/-17_
Concur:
yt-K. P.,EaJbraith, Manager Nuclter ' Safety Engineering Concu h
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Nuclean 64[1s Engineering G.A.Sogej Mapager n
Concur:.
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G. J.vBusselman, Manager Contract Performance Approved:
Ju J9 f.'Nilson~,~Mana9 r 6
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E(ON NUCLEAR COMPANY,Inc.
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IPJ' ORT ANI_ f nilq M G/WDIN, CtPiilNT_5_fgip Ui! QI _TJ45 00y;Pi_NT P_l l A5[
R' AD [ AktilUl_Ly T hi s tec hnica l repor t wa s der i w e.1 t hrouc,b r esea r c h and develo, ent
< ponsored by I n nen TLa lear Company, Inc.
It is t-cinq sutoit tett M
proy,et, 1,y Exxon 'in lear to the lN,KC as part of a technical contribut ion to facilitate safety on.ilyses by licensees of the W/iRC whic h ut ilize lxion
- Nicle,
'a br ica t ed reload f uel or other tec hnic a l servi ( *" provided by i n sor clear for light water power reactors aruf it is true aml (oriett to the Dest of f uon Naclear', k nowled ;e, inf orcu t ion, and bel ie f. The in torn.a t ion c onta ined herein ma y be used 1,y t he IF,NRC in its review of t his report, ani by licensees or applicants Lefore the IMPC weiith are
- r. u s t or,e r s o f Exwn Nuclear in their demonstration of compliance with the USNkC's regalations.
Wit hout deroptin'; f rom t he foregoing, nei t t.er Euon Nuc lear nor an/ persun att1n:] on its behalf:
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Make any warranty, erpress or implitd, with respect to the a((urac y, ( epletenes,, or usef ulne< > of the inforria t ion (ontained in this document, or that the use of any in f orra t ion, appara t us, rie t hod, or prot e,s disc losed in this dacument will not inf ringe privately owned rights; or B.
in sur:e-any liabilities with respec t to the use of, nr for da a.i te re s u l t i r"l f ront the use of, any i n f onaa t ion apparatus,c.ethod, or process disclosed in this datuu.ent.
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i XN-NF-79-21 Revision 1 TABLE OF CONTENTS Section Page
1.0 INTRODUCTION
AND
SUMMARY
1 2.0 BIG ROCK POINT ECCS EVALUATION MODEL.............
5 3.0 BIG ROCK POINT MAPLHGR ANALYSIS RESULTS............
10
4.0 CONCLUSION
S..........................
13
5.0 REFERENCES
15
/
ii XN-NF-79-21 Revision 1 LIST OF TABLES Table Page l.1 Big Rock Point MAPLHGR Summary For G Reload Fuel Types..
3 3.1 Big Rock Point MAPLHGR Versus Fuel Type and Burnup.
12 2
LIST OF FIGURES Figure Page 1.1 MAPLHGR's For G Reload Fuel Types Big Rock Point.
4 2.1 Big Rock Point Blowdown Nodalization Diagram..
8 2.2 Hot Channel Nodalization Diagram....
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_..._ XN-NF-79-21 Revision 1
1.0 INTRODUCTION
AND
SUMMARY
This document presents the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits resulting from a LOCA analysis performed for the Big Rock Point reactor with the Exxon Nuclear Company Non-Jet Pump Boiling Water Reactor (ENC NJP-BWR) ECCS evaluation model.
These results obtained with an Exxon Nuclear Company integrated ECCS model replace those reported previously from the ENC Non-Jet pump BWR Fuel Heatup model which used NSSS vendor input.(I) Details of the ENC NJP-BWR evaluation model and the Big Rock Point system representation are described in References 2, 3, 4, and 5.
The evaluation model was used to perform LOCA-ECCS calculations for a spectrum of break locations, break sizes, and break configurations for Big Rock Point.
The results of these break spectrum calculations are reported in Reference 5.
The worst or limiting 2
break was identified in the spectrum analysis as the 0.375 ft split break between a recirculation pump and the downstream butterfly valve.
For this identified limiting break, heatup analyses have subsequently been performed for ENC reload G, Gl-U, and G3/G4 fuel types. The results of the heatup analyses are allowed MAPLHGR limits for each fuel type versus exposure.
The allowed limits are presented in Table 1.1, and Figure 1.1 for the three fuel types.
Possible axial effects on the midplane MAPLHGR limits were examined, and the midplane MAPLHGR limits were determined to be applicable over the length of the core.
Based on the Big Rock Point analysis results using the ENC NJP-BWR ECCS evaluation model, reactor operation of Big Rock Point within the limits defined by Table 1.1, and Figure 1.1 will assure conformance with 10CFR50.46
__ XN-NF-79-21 Revision 1 criteria for maximum cladding temperature, metal-water reaction, and hydrogen release.
.__..- _ _ XN-NF-79-21 Revision 1 Table 1.1 Big Rock Point MAPLHGR Summary for G Reload Fuel Types ENC RELOAD-G (MO)
ENC G-lu (ALL UO 1 ENC G3/G4 (ALL UO 1 2
2 Burnup MAPLHGR Burnup MAPLHGR Burnup MAPLHGR MWD /MTM kw/ft MWD /MTM kw/ft MWD /MTM kw/ft 0
0 7.214 425 7.618 1,108 7.953 1,754 7.555 1,813 7.808 4,292 7.573 4,334 7.646 7,065 7.282 7,273 7.496 7,261 7.484 14,239 7.165 15,026 7.561 14,877 7.486 21,049 7.113 23,011 7.776 22,375 7.561 27,153 6.843 29,455 7.638 29,054 7.322 31,077 6.081 36,397 7.122 34,399 6.731 37,999 5.734 36,791 6.694 40,000 5.734
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-_. _ XN-NF-79-21 Revision 1 2.0 BIG ROCK POINT ECCS EVALUATION MODEL The evaluation model used to perform the LOCA-ECCS analysis has been described in detail in the example problem submittal and associated documents.(,,5) The RELAP4-EM program was used to perfonn blowdown and hot channel calculations.
Final temperature and metal-water reaction re-sults came from the HUXY code with the integrated BULGEX calculation for cladding swelling and fuel rod rupture.
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 Loss-of-Coolant Accident (LOCA).
The RELAP4-EM blowdown calculation models the recirculation system including the core fuel assemblies, upper plenum, steam drum, steamlines, feedwater lines, pumps, control rod guide tubes, emergency condensers and the lower plenum.
The RELAP4-EM blowdown analysis begins from steady-state or assumed initial operating conditions and continues through pipe rupture and system decompression through the onset of the emergency coolant flow until rated core spray is calculated.
RELAP4-EM/ HOT CHANNEL is used to calculate power, fluid conditions, and heat transfer coefficients in an individual assembly during blowdown using the time dependent boundary conditions in the upper and lower plenum from a prior RELAP4-EM blowdown calculation The HUXY/BULGEX code calculates temperature in the "Jel rods from the initiation of the break until the core sprays quench all rods in the limiting bundle at the axial peak temperature plane of interest.
HUXY considers the heat transfer at a single elevation of a fuel rod bundle and includes con-duction within the fuel rods, convection from the fuel to the fluid, and
- XN-NF-79-21 Revision 1 radiation from rod to rod and rod to cannister.
The Appendix K Yamanouchi correlation is used to calculate channel quench, and the approved ENC 2DQ model is used to calculate quenching of the inert rods.
Fuel rod rupture and ballooning are performed in the BULGEX portion of the model.
Convective coefficients for heat transfer from the rods to the fluid are obtained fran the RELAP4-EM blowdown or hot channel calculations and are input to HUXY to predict clad temperature during the blowdown portion of the transient.
If the quality of the fluid volume containing the plane of interest exceeds 0.999, radiation between the rods and to the cannister is allowed.
HUXY continues the thermal calculation after rated core spray has been obtained using NRC approved coolant spray heat transfer coefficients through the determination of the peak cladding temperature and rod quench.
For the lixll BWR bundle of the Big Rock Point plant, the NRC approved convective heat transfer coefficients (0) are applied af ter rated spray and before quenching.
The BULGEX code, which is an integral part of the HUXY code, calculates clad swelling and fuel rod rupture behavior during the heatup calculation.
For the RELAP4-EM system blowdown analyses, the Big Rock Point plant is modeled as in the example problem (2) using 35 volumes, 46 junctions and 26 heat slabs.
The blowdown nodalization diagram for large breaks is given by Figure 2.1.
Phase separation volumes are input for the steam drum, the emergency condenser, and the downcomers from the steam drum to the pumps.
The blowdown analysis begins from the assumed initial operating con-ditions of 102% of rated reactor power and continues through pipe rupture and system decompression through the onset of the emergency coolant flow until rated core spray is calculated.
..... _ _. _ _ _. XN-NF-79-21 Revision 1 RELAP4-EM/ HOT CHANNEL is used to calculate fluid conditions and heat transfer coefficients in an individual fuel assembly during blowdown using the time dependent boundary conditions in the upper and lower plenum from an applicable prior blowdown calculation.
The nodalization of the hot assembly (Figure 2.2) is also identical to that represented in the example problem. Hot Channel calculations are performed to consider changes in power distribution, fuel assembly design, and exposure.
The HUXY fuel assembly model also remains unchanged from the example problem.(2)
_ XN-NF-79-21 Revision 1
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.. XN-NF-79-21 Revision 1 3.0 BIG ROCK POINT MAPLHGR ANALYSIS RESULTS Maximum Average Planar Linear Heat Generation Rates (MAPLHGR's) 2 were determined based on the identified worst break (0.375 ft split i
break between a recirculation line pump and the downstream valves).
The calculations were made by performing RELAP4-EM hot channel cal-culations with the desired exposure and power distribution conditions assuming plenum boundary conditions for the limiting break.
HUXY calculations followed which were performed for the core midplane at a re-duced power peaking adjusted to meet 10CFR50.46 criteria.
The MAPLHGR values are based on the power used in the heatup analyses.
The MAPLHGR's were computed for three different Exxon Nuclear Company fuel types, Reload G, Gl-U, and G3/G4 for the appropriate exposure ranges to be experienced in future cycles.
The beginning of life (BOL) hot channel calculation used a hot assembly flow of 36.35 lbm/sec based on steady-state thermal hydraulic calculations for an average total core flow 6
of 9.9x10 lbm/hr and a total axial times iadial peaking of 1.96.
Exposed hot channel runs used the same flow but increased total axial times radial peaking to 2.20.
The radial peaking was 1.40 for all cases.
HUXY calculations were performed using the power, heat transfer co-efficient, and coolant temperature data, from the appropriate hot channel calculation results.
The results of the HUXY calculations for MAPLHGR, axial times radial peaking, local metal-water reaction, and peak cladding temperature (PCT) are given in Table 3.1 as functions of exposure for the three ENC fuel types.
These results were those summarized and given in Table 1.1 and Figure 1.1.
a
, XN-NF-79-21 Revision 1 The MAPLHGR's given in Table 3.1 and Figure 1.1 are based on axial power profiles peaked at the core midplane.
The effect of bottom skewed power profiles was also investigated for Big Rock Point. A hot channel calculation was performed assuming an axial peak 1.25 ft above the bottom of the active core.
Total axial times radial peaking was 2.20 and exposed fuel conditions were assumed.
PCT for the HUXY calculation performed for this case (2.20 axial times radial peaking) was 2103 F and the local metal-water reaction was 3.7%.
RELAP4-EM hot channel results showed significantly better blowdown heat transfer for lower hot assembly regions than for the miaplane.
This is due to lower initial fluid quality near the core inlet and a generally positive transient core flow.
The blowdown heat transfer improvement is sufficient to offset the delayed quenching by the core spray at lower elevations.
The MAPLHGR limits given in Table 1.1 and Figure 1.1 are then applicable over the axial length of the Big Rock 4
Point core.
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_, - XN-NF-79-21 Revision 1 Table 3.1 Big Rock Point MAPLHGR's Versus Fuel Type and Burnup ENC RELOAD-G (Mixed Oxide)
r (GWD/MTM)
(kW/ft) ip (op) 7.1 7.28 1.778 5.6 2200 14.2 7.17 1.749 5.6 2200 21.0 7.11 1.737 4.8 2200 27.1 6.84 1.671 5.3 2200 31.1 6.08 1.485 8.4 2150 38.0 5.73 1.400 6.7 2127 40.0 5.73 1.400 8.5 2028 ENC G-lu (All U01 2
Burnup MAPLHGR F xF MWR PCT a I JGWD/MTM)
(kW/ f t).
(%)
( F) 1.8 7.56 1.797 4.1 2200 4.2 7.57 1.801 3.9 2200 7.3 7.50 1.783 3.8 2200 15.0 7.56 1.798 4.0 2200 23.0 7.78 1.849 4.0 2200 29.5 7.64 1.817 4.0 2200 36.4 7.12 1.694 8.5 2160 ENCG3/G4(AllUO1 2
Burnup MAPLHGR F xF MWR PCT a #
(GWD/MTM)
(kW/ft)
,(M
( F) 0 7.21 1.777 4.3 2200
.4 7.62 1.876 4.0 2200 1.1 7.95 1.959 4.4 2200 1.8 7.81 1.923 3.8 2200 4.3 7.65 1.883 4.0 2200 7.3 7.48 1.843 4.1 2200 14.9 7.49 1.844 4.2 2200 22.4 7.56 1.862 3.8 2200 29.1 7.32 1.803 4.0 2200 34.4 6.73 1.658 9.3 2180 36.8 6.69 1.649 10.0 2180
_ _ _ _ XN-NF-79-21 Revision 1
4.0 CONCLUSION
S A LOCA-ECCS analysis has been performed for the Big Rock Point reactor using the ENC WREM-Based NJP-BWR ECCS Evaluation Model which is in conformance to Appendix K of 10 CFR 50.
The analysis included an example problem (2) followed by the required analysis of a spectrum of both large and small breaks and an investigation of the worst break location.I ) Worst single failure assumptions established by previous analyses were assumed since the reactor system remains unchanged.
The limiting break was identified as the 0.375 ft split break between a recirculation line pump and the downstream valve.
Limiting Maximum o
Averige Planar Linear Heat Generation Rates (MAPLHGR's) based on this break were developed for appropriate ENC fuel types and exposures as given in Tables 1.1 and 3.1 and Figure 1.1.
These limits were determined to apply axially over the core length.
Operation of the Big Rock Point reactor with ENC fuel within the limits defined by Table 1.1 assures that the Big Rock Point Emergency Cooling System will meet the acceptance criteria as presented in 10 CFR 50.46 and Appendix K.
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 when the core geometry is still amenable to cooling.
The hot fuel rod cladding oxidation limit of 17% is not exceeded during or after quenching.
... _.. _ XN-NF-79-21 Revision 1 4.
The system long term cooling capabilities provided for previous cores remains applicable to ENC fuel.
_ _ _ XN-NF-79-21 Revision 1
5.0 REFERENCES
1.
Exxon Nuclear Company, ECCS Analysis for Exxon Nuclear Company G-3 All Uranium No Cobalt Fuel For Big Rock Point [ Including Reanalysis of Reload G and G-lU DesignsJ, XN-NF-76-55, Revision 1, February 1977.
2.
Exxon Nuclear Company, Big Rock Point Example LOCA Analysis Using the Exxon Nuclear Company Non-Jet Pump BWR Evaluation Model, Large Break Example Problem, XN-NF-78-25, July 1978.
3.
Letter from G. F. Owsley (ENC) to D. F. Ross (NRC), dated October 30, 1978 regarding RELAP4-EM/ ENC 28B.
4.
Letter from David A. Bixel (CPCo) to Dennis L. Zieman (NRC)
" Docket 50-155 - License DPR Big Rock Point Plant -
Additional Information Concerning ECCS Analysis," dated February 22, 1979.
5.
Exxon Nuclear Company, Big Rock Point LOCA Analysis Using The ENC NJP-BWR ECCS Evaluation Model, XN-NF-78-53, December 1978.
6.
" Safety Evaluation by the Office of Nuclear Regulation Supporting Amendment No. 10 to Facility License No. DPR-6, Consumers Power Company Big Rock Point Plant, Docket No. 50-155,"
dated June 4, 1976; (Included as enclosure in letter from D. L. Zieman of the NRC to R. B. Sewell of Consumers Power Company, dated June 4, 1976).
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