ML20198N534
| ML20198N534 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 10/31/1997 |
| From: | Richard Anderson NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9711050229 | |
| Download: ML20198N534 (5) | |
Text
'.
Northern States Power Company Prairie Island Nuclear oenerating Plant 1717 Wakonado Dr East Welch. Mennesota $5089 October 31,1997 U S Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GEN 6 RATING PLANT Docket Nos. 50-282 License Nos. DPR-42 50 306 DPR-60 Use of Sandberg Correlation for Generating Cladding Surface Heat Transfer Coefficients This letter is to inform the NRC of an issue discovered during a self assessment of NSP's reload methodology. The issue relates to a discussion in the February 17,1983 NRC Safety Evaluation for NSP Topical Report NSPNAD-8102 A," Prairie Island Nuclear Power Plant Reload Safety Evaluation Methods for Application to PI Units,"
regarding the use of the Sandberg Correlation for generating the fuel cladding heat transfer coefficient. Specifically, the NRC Safety Evaluation implies that the benchmark cases run for the rod ejection analysis used the Sandberg Correlation as outlined in the proceduto described in a January 4,1983 NSP letter, when in fact the benchmark cases did not utilize the Sandberg Correlation.
NSP has conducted a review of the NRC Safety Evaluation and has determined, as
{i described in the attached discussion, that this inconsistency in the NRC Safety s
Evaluation does not alter the NRC's approv' % use the Sandberg correlation for reload analyses.
,bd)\\
In this letter we have made no new Nuclear Regulatory Cornmission commitments.
Please contact Gene Eckholt (612-388-1121)if you have any questions related to this letter.
41 h-lA%
Roger O Anderson Director Nuclear Energy Engineenng 9711050229 971031 PDR ADOCK 05000202 P
PDR J
4 USNRC NORTHERN STATES POWER COMPANY October 31,1997 Page 2 c: Regional Administrator - Region Ill, NRC Senior Resident inspector, NRC NRR Project Manager, NRC J E Silberg
Attachment:
e i + 5. i g t;
\\
l Use of Sandbera Correlation for Generatina Claddina Surface Heat Transfer Coefficients Discussion in Februsry 1982, NSP submitted a Topical Report (Reference 1) to the NRC that described ths methodology for accident analyses that NSP Intended to use for evaluating future reloads and plant modifications. This Topical Report also coniained summaries of benchmark cases to demonstrate that NSP understood "the manner in which the bounding physics parameters have been used in each analyses and the conservatism inherent in the values chosen", and to demonstrate NSP's safety analysis experience and expertise.
For the Rod Ejection accident, this was accomplished by reproducing the results in WCAP 7588, "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods" (Reference 5). Since the WCAP did not list many of the key inputs used in the analysis it was necessary to perform sensitivity studies on selected inputs to determine which value would provide the most meaningful results. Specifically, sensitivity studies were run varying the cladding surface heat transfer coefficient to determine the impact it had on the peak cladding temperature. These studies were discussed in the Topical Report. From these studies, specific heat transfer coefficient values were selected that resulted in the benchmark cases generating temperature profiles consistent with those in WCAP-7588. The Sandberg Correlation was not used to generate the benchmark heat transfer coefficient values. The cladding surface heat transfer coefficient values used in the benchmarks were never intended to be viewed as methodology requirements for future reload analyses.
On July 21 1982, NSP made a presentation to the NRC regarding the Topical Report and qualifications of NSP personnel that would be performing the analyses. In the summary of the meeting (Reference 2), the NRC requested additionalinformation including "A description of how the TOODEE2 Code is used in the reload analysis."
The response to this request took the form of Appendix D in Revision 1 of the Topical Report (Reference 3) and was submitted to the NRC in December of 1982. This appendix described how NSP intended to generate the cladding heat transfer coefficient based on the Saridberg correlation for future reload analyses.
Following the submittal of Revision 1 of the Topical Report, NSP sent a letter dated January 4,1983 with supplementalinformation concerning the use of the TOODEE2 code (Reference 6). This letter provided more details on the assumptions that would be used in running the TOODEE2 code for future reloads including how the Sandberg correlation would be used to generate the cladding heat transfer coefficient.
\\
Attachment Octobu 31,1997 Page 2 The NRC issued their Safety Evaluation (Reference 4) for the Topical Report in February 1983. The NRC Safety Evaluation stated that NSP intended to use the Topical Reports as guides to analyze future reloads and operations. It also acknowledged that the benchmarks in the Topical Report provided a summary of NSP's calculational experience and provides a good deal of discussion regarding the acceptability of the codes in performing reload analyses. Based on these statements, as well as discussions with individuals involved with the generation of the Topical Report who were present et the July 21,1982 meeting, it has been concluded that the NRC understood that the purpose of the benchmark cases in the Topical was to demonstrate NSP's experience in performing transient analyses and not to specify input assumptions (e.g. the value of the cladding surface heat transfer coefficient). The methodology requirements for reload analyses were specified in the Topical Report
" Accident Analysis' subsections as well as in the Appendices.
The second full paragraph on page 24 of the NRC Safety Evaluation describes the initial, bounding, and transient conditions used in the TOODEE2 code as well as the method for calculating the cladding surface heat transfer coefficient using the Sandberg correlation that NSP intended to use for future reloads. This paragraph includes a discussion that implies both the Locked Rotor and Rod Ejection benchmark analyses used this method. As stated above, the Rod Ejection benchmark cases used cladding surface heat transfer coefficients that provided temperature profiles consistent with those in WCAP 7588 and were not based on the sandberg correlation. A review of relevant correspondence and meeting minutes found no evidence that NSP ever stated that the Rod Ejection benchmark cases in the Topical (Figures 3.15 2 and 3.15-4) used the Sandberg correlation to generate the cladding heat transfer coefficient.
The NRC Safety Evaluation goes on to state the NRC reviewed NSP's methodology of using the TOODEE2 code for non LOCA transients, including events similar to Locked Rotor and Rod Ejection, and found it to be acceptable. As stated above, the methodology for genereting the value of the cladding heat transfer coefficient is defined in Appendix D of the Topical and not defined in the benchmark cases.
Summary The NRC Safety Evaluation for NSP Topical Report NSPNAD-8102-A contains a discussion that implies that the cladding heat transfer coefficient in the Rod Ejection benchmark cases were generated based on the Sandberg correlation, when in fact they were not. NSP has concluded that this discussion does not alter the NRC's approval to use the Sandberg correlation, as described in Appendix D of the Topical, for generating the cladding heat transfer coefficient for use in the TOODEE2 code for non-LOCA transients including the Rod Ejection transient.
f,'T Wh Da
- \\
t Attachment October 31,1997 Page 3 References
- 1. NSPNAD 8102," Prairie Island Nuclear Power Plant Reload Safety Evaluation Methods for Application to P1 Units", December 1981.
- 2. " Summary of July 21.1982 Meeting with NSP Regarding Fuel Reload Methodology for Prairie Island Unit Nos.1 and 2"; Dominic C. Dilanni, Project Manager Operating Reactors Branch #3 Division of Licensing.
- 3. NSPNAD-8102, Revision 1, *Prrsirie Island Nuclear Power Plant Reload Safety Evaluation Methods for Applicttion to PI Units", December 1982
- 4. Letter from: Robert A. Clark, Chief Operating Reactors Branch #3 Division of Licensing; to: D. M. Musolf, Nuclear Support Services Department Northern States Power; Dated February 17,1983.
- 5. WCAP 7588 Revision 1 A,"An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods",
January 1975.
- 6. Letter from: D. M. Musolf, Manager Nuclear Support Service; to: Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulator Commission, Dated January 4, 1983;
Subject:
" Supplement Information for the Review of the NSP Safety Evaluation Methods Topical".
aw n
____.-..__._d