ML20034D469
| ML20034D469 | |
| Person / Time | |
|---|---|
| Site: | Columbia, Washington Public Power Supply System  |
| Issue date: | 08/28/1992 |
| From: | Rosalyn Jones Office of Nuclear Reactor Regulation |
| To: | Yung Y WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| References | |
| TAC-M79498, NUDOCS 9209220471 | |
| Download: ML20034D469 (2) | |
Text
r August 28, 1992 Fir.
Y. Y.
Yung, Chairman JAPAE vr naintenance troup Washington Public Power Supply System Post Office Box 968 Richland, WA 99352
Dear Mr. Yung:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION ON VIP"E-01 MOD-02 DOCUMENTATION EPRI NP-2511-CCM, VIPRE-10: A THERMAL-HYDRAULIC ANALYSIS CODE FOR REACTOR CORES (TAC No. M79498)
References:
1.
"VIPRE-10: A Thermal-hydraulic Analysis Code for Reactor Cores, " Volumes 1 - 4, EPRI-2511-CCM-A, Revision 3, August 1989.
2.
Letter from Y. Y. Yung (VMG) to USNRC,
" Notification of Release and Recuest fnr NRC Review of VIPRE-01 MOD-02," February 28, 1990.
3.
Letter from Y. Y. Yung (VMG) to USNRC, "VIFRE-01 Error / Change Log," February 26, 1991.
4.
Letter from R.
C.
Jones, USNRC, to Y.
Y.
Yung (VMG), " Request for Additional Information on VIPRE-01 MOD-02," September 3.
1991.
5.
Letter from Y. Y. Yung (VMG) to R.
C.
Jones (NRC), March 16, 1992,
" Response-to Request for Additional Information on VIPRE-01 MOD-02 Documentation EPRI NP-2511-CCM, VIPRE-01: A t
Thermal-Hydraulic Analysis Code For Reactor Cores (TAC No. M79498)"
The Reactor Systems Branch of the NRC and International Technical Services, Incorporated (ITS) have reviewed your response (Ref. 5) i to the Request for Additional Information (Ref. 4).
We would like you to 1) expand on the first question in reference 4 which
+
is modified with minor changes below and 2) supply the reference requested below.
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Mr.
Y. Y. Yung, Chairman :
1.
For each of the changes listed in Table 1 of Reference 2, quantify for each code thermal-hydraulic"parametef and the DNBR, the magnitude and direction of the change with respect to BWR and PWR applications.
If the impact of change is-transient dependent, identify the affected variables, delineate the conditions which result in the extreme cases, and provide the magnitude of the differences.
For example, the discussion of changes 125 and 128 must cover transient conditions and must delineate the worst case transient.
Basically this question is the same as the last time for Question Number 1, with minor changes.
2.
Provide Reference 45 (Correlation of Critical Heat Flux Data for Application to Boiling Water Reactor Conditions, Hench, J.E.
and J.C.
Gillis, EPRI-1898, June 1981) on page 4-4 of Volume 1 (Mathematical Modeling (Revision 2)).
i Please contact Harry Balukjian of my staff at 301-504-2862 if there are any questions,._
~ ~..,.. d b^i 3.Si C. d;GL Robert C. Jones, Chief Reactor Systems Branch Division of Systems Technology Office of Nuclear Reactor Regulation cc: H.Komoriya DISTRIBUTION:
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