ML18094A389
| ML18094A389 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 04/27/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML18094A388 | List: |
| References | |
| NUDOCS 8905030569 | |
| Download: ML18094A389 (3) | |
Text
UNITED STATES NUCLEAR REGULATORY COMM'rsslON WASHINGTON, D. C. 20555 ENCLOSURE 1 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REVIEW OF LOCA/ECCS REANALYSIS PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM UNITS 1 AND 2 DOCKET NOS. 5d-272 AND 50-311
1.0 INTRODUCTION
By letter dated March 31, 1989 {Ref. 1), Public Service Electric and Gas Company submitted a revised large break LOCA analysis for Salem Generating Station, Units 1 and 2. The analysis is in support of Salem Unit 2 Cycle 5 operation and also in support of the Vantage 5H fuel License Change Request **
which is applicable to the Salem Unit 1 Cycle 9 core reload. The reanalysis for Unit 2 is responsive to NRC Generic Letter 86-16 which requires a reevaluation of the large break LOCA spectrum whenever plant changes are made which may affe.ct the results of the original license model to confirm that the plant meets the applicable criteria of 10 CFR 50.46(b) based on the current plant configuration.
2.0 EVALUATION The required reanalysis was prompted by two identified changes in plant configuration. The discovery of defective tubes in two Salem Unit 2 steam generators resulted in 2.7 percent of the Unit 2 steam generator tubes requiring plugging. Also during the fourth refueling outage, a burnable poison rodlet assembly hold down nut, a locking weld pin, and a hand held ga1m1a measurement probe with cable connector were inadvertently dropped into the reactor cavity. Unsuccessful efforts to retrieve these items resulted in a decision to evaluate the objects as loose parts within the reactor cooling system (RCS).
The reanalysis was performed, in part, to determine the potential effect of the above conditions on peak clad temperature {PCT) during a large break LOCA.
Reference 1 includes two attachments which document the analyses applicable to the two plant changes. Attachment 1 (Ref. 2) provides the results of an analysis assuming 10 percent steam generator tube plugging in either Unit 1 or Unit 2 and Attachment 2 (Ref. 3) provides an evaluation of the effect of postulated loose parts on the new large break LOCA analysis results for Salem Unit 2. The submittal was evaluated by the staff in two parts as follows.
2.1 LOCA analysis assuming 10 percent steam generator tube plugging Reference 2 presented the results of a Salem large break loss-of-coolant accident analysis performed using the 1981 Westinghouse Emergency Core Cooling System Evaluation Model with the BASH code. The assumption of 10 percent tube plugging for all steam generators bounds the actual 2.7 percent plugging performed for Salem Unit 2 and would apply as well to Unit 1 1f a similar tube plugging effort should be required. Other analysis assumptions were (a) an increase in the total peaking factor to 2.40, (b) an analysis of the more limiting fuel type {17x17 standard fuel) in either unit, (c) an increase in the hot channel enthalpy rise factor to 1.60, (d) a degradation in Thermal Design Flow of 1 percent, (e) a limited degradation in Safety Injection performance to reflect the current plant condition, (f) an increase in diesel generator/safety injection delay time from 25 seconds to 30 seconds, and (g) a lowering of low pressurizer pressure reactor trip and low low pressurizer pressure Safety Injection setpoints. The Westinghouse 1981 Evaluation Model (EM) was used in conjunction with the staff approved BASH methodology.
The staff Safety Evaluation Report on the use of the 1981 EM with BASH (Ref. 4) imposed restrictions which were conformed to in the reanalysis. Those restrictions applicable to Salem were (a) separate analyses with minimum and maximum safety injection, and (b) confirmatory analyses to demonstrate that the cosine power shape is limiting and is the appropriate power shape to use for licensing calculations. The result of the reanalysis showed a Peak Clad Temperature {PCT) of 2019°F which is below the allowable limit. Therefore the reanalysis documented in Reference 2 is acceptable.
2.2 Effect of loose parts on LOCA analysis Reference 3 took into consideration 'the additional effect of loos~~parts on the new LOCA analysis for Salem Unit 2.
The assumption that all three of the identified objects remain in the hottest core subchannel is suitably conservative since this has the greatest potential to affect the PCT calculation.
The estimated subchannel blockage of 36 percent at the grid elevations created a PCT increase of about 25°F.
The maximum PCT for any grid elevation in the hot channel was 2007°F.
The grid location is not the critical location for maximum PCT since the local power is less and heat transfer is higher in the region of the grids. Consequently the resulting clad temperature is not limiting compared to the BASH limit for tube plugging (independent of grid location) of 2091°F.
The combined effect of tube plugging modifications and the potential blockage effect of loose parts would result in a PCT less than the 10 CFR S0.46(b) limit.
2.3 Transition core considerations The transition core under consideration for Salem Unit 2 is from the use of 17x17 standard (STD) fuel to VANTAGE 5 Hybrid (VANTAGE SH) fuel without Intermediate Flow Mixers (!FM).
The generic aspects of the VANTAGE SH fuel assembly are discussed in WCAP-10444 {Ref. S). Considerations relative to the VANTAGE SH fuel assembly included the staff Safety Evaluation (Ref. 6) which requires that an assessment of a transition core must include a demonstration whether a greater peak clad temperature than that calculated for a complete VANTAGE SH configuration can occur for the transition configuration. The conclusion by the staff was that the transition core penalty specified in WCAP-10444 results in a conservative estimation of the large break LOCA PCT and is acceptable.
The greatest penalty for any transition core identified in WCAP-10444 is less than 100°F which is within the margin for the Salem 2 reanalysis result reported in the licensee's submittal.
". The staff concludes that all substantial aspects of the plant reconfiguration have been addressed in an acceptable manner and the requirements for a LOCA reanalysis for Salem Units 1 and 2 have been satisfied.
3.0 CONCLUSION
S Based on our review of the information presented in the licensee's letter dated March 31, 1989 and attachments thereto, the staff has concluded that the licensee has provided sufficient information for the staff to conclude that (a) the changes in plant configuration resulting from steam generator tube plugging and loose parts in the reactor coolant system (RCS) will not adversely affect the safety margin of Salem Units 1 and 2, (b) those aspects of the transition core have been adequately considered and the results are acceptable, and (c) the requirement for a formal reanalysis identified in the temporary exemption from 10 CFR 50.46{a)(l)(i) has been satisfied.
In addition, those aspects relative to a transition core have been acceptably addressed.
4.0 REFERENCES
- 1.
- 2.
- 3.
- 4.
- 5.
- 6.
Letter S. E. Miltenberger (PSEG) to Document Control Desk (USNRC) dated:
March 31, 1989, subject: "Revised Large Break LOCA Analysi- Salem Generating StatiOtt *unit Nos. 1 and 2" {with attachments).
WCAP-12192, Salem Units 1 and 2 10 Percent Tube Plugging Large Break LOCA BASH Analysis, Westinghouse Electric Corporation, March 1989, (attachment to Reference 1).
SECL-88-547, Rev. 1, LOCA Bases - Large Break LOCA - FSAR Chapter 15.4.1, March 8, 1989.
WCAP-10266, Rev. 2 with Addenda (Proprietary) "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code," November 13, 1989.
WCAP-10444{P){A) Addendum 2, VANTAGE SH Fuel Assembly Westinghouse Electric Corporation, April 1988 (with SE attached).
Letter A.C. Thadani, NRR, to R.A. Wieseman WEC, "Acceptan~e for Referencing of Topical Report WCAP-10444(P~ Addendum 2 "VANTAGE 5H Fuel Assembly," November 1, 1988.