ML20070S998
| ML20070S998 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 01/31/1983 |
| From: | Musolf D NORTHERN STATES POWER CO. |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8302080144 | |
| Download: ML20070S998 (3) | |
Text
r Northem States Power Company 414 Nicollet Mall Minneapohs. Minnesota 55401 Telephone (612) 330-5500 January 31, 1983 Director Office of Nuclear Reactor Regulation U S Nuclear Regulatory Commission Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Supplement Information for the Review of the NSP Safety Evaluation Methods Topical - COBRA-IIIC/MIT - Correction Please replace the attachment to the January 28, 1983 letter with the enclosed attachment.
Data was missing from the Locked Rotor Analysis Table on page one.
D M Mws D M Musolf, Manager Nuclear Support Servi s DMM/TMP/ dab cc Regional Admin-III, NRC NRR Project Mgr, NRC NRC Resident Inspector l
G Charnoff Attachment I
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l 8302080144 830131 PDR ADOCK 05000282 P
.e Not all of the local parameters can be corrected to the FSAR value since some of these are not shown. However, by correcting the few variables
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shown above, the MONBR results show better agreement with the FSAR indicating the deviations in the original calculation of MONBR were due to the DYN0DE-P input and not in the COBRA IIIC/MIT code. Note that in the Slow Rod Withdrawal case the initial DNBR calculations in the FSAR disagree by approximatly 0.07 with both that calculated by COBRA IIIC/MIT and also with the initial values from all other transients shown in the FSAR. This indicates a probable error in Figure 14.1-9 of the FSAR. If this bias of -0.07 is applied to the NSP revised calculation, the MDNBR becomes 1.464 which is in better agreement with the FSAR value of 1.36.
Also, the FSAR calculations were probably run with a single channel model.
Figure 3.2-8 of reference 1 shows a difference of,approximately
-0.07 between the single channel model and the 1/8 assembly model MDNBR.
If this bias is applied to the revised calculation, it brings the MDNBR value down to 1.394 which is in good agreement with the FSAR number.
REFERENCES 1.
Prairie Island Nuclear Power Plant " Reload Safety Evaluation Methods for Application to PI Units," NSPNAD-8102P Rev 1, December 1, 1982.
2.
Northern States Power Company, Prairie Island Nuclear Power Plant, Final Safety Analysis Report.
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1 I
I RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION ON NSPNAD-8102P The following Locked Rotor and the Slow Rod Withdrawal transients were rerun, with respect to MDNBR, using input parameters, to COBRA IIIC/MIT, which better represent the local conditions calculated in the FSAR. The following results were obtained using an 1/8 assembly COBRA model(1) run in a psuedo steady state mode at the point of minimum DNBR in the transient.
LOCKED ROTOR ANALYSIS Reference (1)
Revised Calculation Calculation FSAR(2j Pressure (psia) 2410 2740 2740 2
Heat Flux (MBtu/hr ft )
, 0.2317 0.2317 Inlet Temperature (*F) 540.0 540.0 2
Inlet Flow (M1bm/hr ft )
1.124 1.124 MDNBR 1.068 1.375 1.30 SLOW ROD WITHDRAWAL ANALYSIS l
Reference (I)
Revised Calculation Calculation FSAR(2)
Pressure (psia) 2326 2355 2355 2
Heat Flux (MBtu/hr ft )
0.2821 0.3264 0.3264 Inlet Temperature (*F) 541.7 541.7 2
Inlet Flow (Mlbm/hr ft )
2.400 2.400 MDNBR 1.691 1.534 1.36 Initial DNBR 1.891 1.891 1.82 l
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