ML20100J316
| ML20100J316 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 04/04/1985 |
| From: | Andrews R OMAHA PUBLIC POWER DISTRICT |
| To: | John Miller Office of Nuclear Reactor Regulation |
| References | |
| LIC-85-141, NUDOCS 8504100447 | |
| Download: ML20100J316 (9) | |
Text
-
Omaha Public Power District 1623 Harney Omaha, Nebraska 68102 402/536-4000 April 4,1985 LIC-85-141 i
Mr. James R. Miller, Chief Office of Nuclear Reactor Regulation Division of Licensing U. S. Nuclear Regulatory Commission Washington, D. C.
20555 k
References:
- 1) Docket No. 50-285
- 2) Letter from OPPD (R. L. Andrews) to NRC (James R. Miller) dated March 25, 1985, LIC-85-122
Dear Mr. Miller:
Error in LOCA-ECCS Analysis for Fort Calhoun Station On March 25, 1985, Omaha Public Power District provided Reference 2, its initial response, to the error in Exxon Nuclear Company's (ENC) LOCA coding. Subsequently, a meeting was held on Friday, March 29, with the NRC staff, ENC representatives, and utilities using ENC methodology for 3
LOCA analysis in attendance. Based on discussions during this meeting, the Omaha Public Power District is providing the following infomation relative to the ENC LOCA analysis applicable to the Fort Calhoun Station.
A description of models used by ENC in the Fort Calhoun LOCA analysis for Cycle 8 and a description of models used in the reanalysis after the ENC coding misformulation are contained in Table 1.
Table 2 lists the ENC models and their appropriate references.
Non-uniform Fort Calhoun specific axial power distributions were generated for use in the Fort Calhoun LOCA analysis performed by ENC as discussed in Section 14.15 of the USAR. These axial gower distributions were used by ENC in the generation of the enclosed Fgi versus core height curve.
T The FO value of 2.53 is equivalent to a Peak Linear Heat Generation Rate (PLHGR) limit of 15.22 kw/ft. The PLHGR limit is a Limiting Condition 7
of Operation (LCO) and is included as Technical Specification Figure 2-5.
This LCO is monitored in accordance with Technical Specification 2.10.4(1). The linear heat rate limit is nomally monitored by the incore detectors which are 40 cm long and their midpoints are nominally located at 20, 40, 60, and 80 percent of core height. This incore detector moni-T toring system assures that the FQ limit is maintained at or below 80 8504100447 850404 PDR ADOCK 05000295 800f P
PDR 45 5t24 Employmen h quet Opportunity
l Mr. James R. Miller LIC-85-141 Page Two percent of core height. The LCO for FxyT (Technical Specification Figure 2-9) in conjunction with the full power DNBR LC0 axial shape limit (Tech-nical Specification Figure 2-7) preclude the total peaking factor above the 80 percent core height level from reaching the limits prescribed in the enclosed figure. As a nomal part of the generation or verification of the above-mentioned LCO, calculations are performed for reload cores to verify that total peaking above the 80 percent core height will not exceed the limits defined in the enclosed figure.
Based on Reference 2 and the enclosed information, the District finds that Fort Calhoun Station remains in compliance with 10 CFR.50.46.
Sincerely, R. L. Andrews /ev1 Division Manager 4
Nuclear Production RLA/JDK/1p Enclosures cc:
LeBoeuf, Lamb, Leiby & MacRae 1333 New Hampshire Avenue, N. W.
Washington, D. C.
20036 Mr. E. G. Tourigny, NRC Project Manager Mr. L. A. Yandell, NRC Senior Resident Inspector i
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- Table 1 Fort Calhoun LOCA/ECCS Model Summary Current
- _ Analysis 1)
Fission-Gas Release Model WREM WREM 2)
Stored Energy Model WREM WREM 3)
Blowdown Model WREM WREM
- 4)
Containment Model WREM WREM 5)
Clad Swelling and Rupture Model
.EXEM-NOREG EXEM-NUREG
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6)
Reflood Mode 1~
f a)
Carryout and Quench Correlation WREM WREM b)
Downcomer/ Upper Plenum Leakage WREM:
WREM:
No Leakage No Leakage c)
Break Model CD = 1.0 CD = 1.0 Guillotine Guillotine d)
Core Outlet Enthalpy Model EXEM EXEM e)
Z-Equivalent Model 0FF 0FF
- 7) -heatup Model L[
's)
Steam Cooling Model EXEM EXEM b
Heat Transfer Correlation WREM WREM i;
c Mixing Vane Multiplier 1.0 1.0
/
d Local Faaking Multiplier 1.045 1.0
.e Z-Equivalent Model WREM WREM
!f Radiation Model ON ON 4
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"8)
Documentation of Results OSAR 14.15 Letter i
Loss of LIC-85-122
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Current analysis performed to correct error in code T00DEE2.
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.._z Table 2 Exxon Nuclear Company ECCS Models Model Reference 1.
Fission Gas Release Model a.
SAPEX WREM 1
.b.
GAPEX with Uncertainties WREM-II 1, 6 c.
RODEX2 EXEM 13 2.
Stored Energy Model a.
GAPEX WREM 1
b.
R00EX2 EXEM 13 RODEX2igRELAP4 EXEM 11c, 13 c.
3.
Blowdown Model WREM
- 3. 5, 7 4.
Containment Pressure Model a.
Dry Containment WREM 3
b.
Ice Condenser Containment WREM-II 6.-14 5.
Clad Swelling and Rupture Model a.
Exxon Model WREM 3, 4 b.
Revised Exxon Model WREM-II 3,4,6 c.
Exxon Model including NUREG-0630 EXEM
. 10 6.
Reflood Model a.
RELAP4' WREM 3
b.
REFLEX EXEM 8
c.
Carryout and Quench Correlations 1
~15x15 FLECHT WREM
- 2 2
17x17 FLECHT EXEM 11a 3
15:15/17x17 FLECHT EXEM 11b d.
Downcomsr/ Upper Plenum Leakage EXEM lla e.
Break Model 1)
Split Break EXEM 11a
'2)
Guillotine Break EXEM' 11a
.f.
Core Outlet Enthalpy Model EXEM 113 g.
Z-Equivalent Model 1
WREM.
WREM 3
2 EXEM EXEM 11d
Table 2(Continued) 7.
Fuel Rod Heatup Model a.
T00DEE2 WREM 3
b.
Steam Cooling Model 1
WREN WREM 3, 9 2
WREM-II WREM-II 6, 9 3
EXEM EXEM 11 c.
Heat Transfer Correlation 1
15x15 WREM 3
2 15x15 WREM-II 6
3 17x17 EXEM lla 4
15x15/17x17 EXEM 11b d.
Mixin'g Vane HTC Multipliers 1)
Off WREM 3
2)
EXEM EXEM 11a e.
Local Peaking HTC Multipliers EXEM 11 1)
Off WREM 3.
I 2)
EXEM EXEM lla 3) 0.C. Cook 2 EXEM 16
- f.
2-Equivalent Model 1)
WREM WREM 3
2)
EXEM EXEM 11d 9
Radiation Model 1)
WREM WREM 3
2)
WREM Expanded New Geometries EXEM 11d 8.
Core Wide Metal-Water Reaction WREM 15 4
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Table 2 (Con 51nued)
REFERENCES XN-73-25, "GAPEXX: A Computer Program for Predicting Pellet-to-Cladding 1.
Heat Transfer Coefficients," Exxon Nuclear Company Inc., Richland, WA 99352, August 1973.
XN-75-19, " Carryout Rate Fraction Correlation for Pressurized Water 2.
Reactors," Exxon Nuclear Company, Inc., Richland, WA 99352; (a) March 24, 1975; (b) Supplement 1, " Statistical Evaluation of the Carryout Rate Fraction," June -1975.
3.
XN-75-41, " Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model," Exxon Nuclear Company, Inc., Richland, WA 99352.
a.
Volume 1,' July 25, 1975.
b.
Volum'e 2, "Model Justification," August 1, 1975.
Supplement 1. "Further Definitions and Justifications to Reflood 4
c.
Heat Transfer Models," August 14, 1975.
d.
Supplement 2. " Supplementary Information Related to Blowdown and Reflood Analysis," August 14, 1975.
Supplement 3, " Supplementary Information Related to Blowdown and e.
Heatup Analysis," August 16, 1975.
f.
Supplement 4. " Supplementary Information Related to Blowdown and Heatup Analysis " August 20, 1975.
g.
Supplement 5 " Supplementary Information Related to Blowdown and Heatup Analysis," October 3,1975.
h.
Supplement 6. " Supplementary Information Related to Blowdown and Heatup Analysis," October 27, 1975.
i.
Supplement 7 " Supplementary Information " November 9, 1975.
J.-
Volume II Appendix A and B, "3-Loop Westin'ghouse Sample Problem,"
August 1, 1975.
k.
Volume II Appendix C " Yankee Rowe Example Problem," August 22, 1975.
j 1.
Volume II Appendix D, "3-Loop Westinghouse Large Break Example
' Problem (Using September 26, 1975 Model)," October'2, 1975.
m.
Volume III Revision 2, "Small Break Model," August 20, 1975.
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Table 2 (Continued) 0
-REFERENCES _(Continued) 4.
XN-75-6 " Flow Blockage Model for LOCA Analysis," Exxon Nuclear Company, Inc.,RIchland,WA99352, April 1,1975.
5.
XN-75-43, " Core Physics Methods and Data Used as Input to LOCA Analysis,"
Exxon Nuclear Company, Inc., Richland, WA 99352, August 1975.
6.
XN-76-27(A), " Exxon Nuclear Company WREM-Based Generic PWR ECCS Evalu-ation Model Update ENC WREM-II," Exxon Nuclear Company, Inc., Richland, WA 99352; (a) March 1977; (b) Supplement 1(A), " Supplementary Infor-mation Relating to," March 1977; and (c) Supplement 2(A), " Supplementary Information Relating to," March 1977.
7.
XN-76-44,'" Revised Nucleate Boiling Lockout for ENC WREM-Based ECCS Evaluation Model," Exxon Nuclear Company Inc., Richland, WA 99352, September 1976.
B.
XN-NF-78-30(A), " Exxon Nuclear Company WREM-Based Generic PWR ECCS Model Updates ENC WREM-IIA," Exxon Nuclear Company, Inc., Richland, WA 99,352, May 1979.
9.
Letter, G.F. Owsley (Exxon Nuclear Company) to D.F. Ross (USNRC).
Subject:
T00DEE2 Updates; Letter No. GF0:077:80 dated April 1, 1980.
- 10. XN-NF-82-07(P)(A), Rev. 1, " Exxon Nuclear Company ECCS Cladding Swelling
_and Rupture Model," Exxor. Nuclear Cospany, Inc., Richland, WA 99352, Novereber 1982.
- 11. XN-NF-82-20(P), " Exxon Nuclear Company Evaluation Model EXEM/PWR ECCS Model Updates," Exxon Nuclear Company, Inc., Richland, WA 99352; (a)
Revision 1, August 1982; (b) Revision 1, Supp.1. " Revised FLECHT-Based Reflood Carryover and Heat Transfer Correlations,"~ June 1983;
. (c) Revision 1. Supp. 2(A), February 1985; (d) Revision 1 Supp. 3,
" Response to NRC Request for Additional Information " Draft; (e) Revision
- 1. Supp. 4(A), " Adjustments to FLECHT-Based Heat Transfer Correlations,"
July 1984.
- 12. XN-NF-82-49(P), " Exxon Nuclear Company Evaluation Model - EXEM/PWR Small Break Model," Exxon Nuclear Company, Inc., Richland, WA 99352; (a) June 1982; (b) Supp. 1. " Supplement 1: Renonses to NRC Questions,"
March 1985.
- 13. XN-NF-81-58(P)(A), Revision 2, Supps.1 & 2. "RODEX2 Fuel Rod Thermal Response Evaluation Model," Exxon Nuclear Company, Inc., Richland, WA-99352, March 1984,
- 14. XN-CC-39, Rev. 1. "ICECON: A Computer Program Used to Calculate Containment Back Pressure for LOCA Analysis (Including Ice Condenser Plants),"
Exxon Nuclear Company, Inc., Richland, WA 99352, November 1978.
1 Table 2 (Continued)
\\
REFERENCES (Continued)
- 15. XN-CC-36, " Exxon Nuclear Procedure for Calculating Core-Wide Metal-Water Reaction During a Loss-of-Coolant Accident," Exxon Nuclear Company, Inc., Richland, WA 99352 December 1975.
- 16. Letter, J.C. Chandler (Exxon Nuclear Company) to H.R. Denton (USNRC),
Subject:
Local Peaking Multiplier for Reflood Heat Transfer Coefficient; Letter No JCC:076:84 dated May 7, 1984.
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