ML17311A329
| ML17311A329 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 10/07/1994 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17311A328 | List: |
| References | |
| NUDOCS 9410130136 | |
| Download: ML17311A329 (6) | |
Text
~PS RE00 0
Cy o
I v
4 0
+>>**+
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 9410i3013b 941007 PDR ADOCK 05000528' PDR SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.
83 TO FACILITY OPERATING LICENSE NO. NPF-41 MENDMENT NO. 70 TO FACILITY OPERATING LICENSE NO. NPF-51 AND AMENDMENT NO. 55 TO FACILITY OPERATING LICENSE NO. NPF-74 ARIZONA PUBLIC SERVICE COMPANY ET AL.
PALO VERDE NUCLEAR GENERATING STATION UNIT NOS.
1 2
AND 3 DOCKET NOS.
STN 50-528 STN 50-529 AND STN 50-530
- 1. 0 INTRODUCTION By letter dated August 18, 1994, the Arizona Public Service.Company (APS or the licensee) submitted a request for changes to the Technical Specifications (TS) for the Palo Verde Nuclear Generating Station, Units 1, 2, and 3
(Appendix A to Facility Operating License Nos.
NPF-41, NPF-51, and NPF-74, respectively).
The Arizona Public Service Company submitted this request on behalf of itself., the Salt River Project Agricultural Improvement and Power District, Southern California Edison
- Company, El Paso Electric 'Company.,
Public Service Company of New Mexico, Los Angeles Department of Mater and Power,,
and Southern California Public Power, Authority.
The proposed changes would revise TS 6.9. 1. 10 to add, the analytical method supplement entitled "Calculative Methods. for-the CE Large Break LOCA Evaluation Model,for the Analysis of CE and W Designed NSSS,"
CENPD-132, Supplement 3-P-A, dated June 1985.
This TS contains the list of analytical methods used to determine the PVNGS -core operating 1-imits.
Additionally, APS proposed to delete the existing references to earlier versions of CENPD-132, and the associated approval
- letters, supplanted by Supplement 3-P-A, and reletter the remaining list.
Technical Specification 6.9. 1. 10 lists'he analytical
- methods, previously reviewed and approved by the staff, that are used to determine the core operating limits for the PVNGS plants.
Plant operation is limited in accordance with the values of cycle-specific parameter limits that are established using these NRC-approved analytical methods.
The analytical method, proposed for addition to TS Section 6.9. 1.10 supplants those currently used for Specification 3.2. 1, Linear Heat Rate, for large break loss-of-coolant accident (LOCA) analysis.
- 2. 0 EVALUATION The large break LOCA analytical methods currently listed in TS 6.9. 1. 10.e, f, and g assume up to 1100 plugged steam generator tubes.
The licensee anticipates that this steam generator tubes plugging assumption will soon be
Il I
tI~"
exceeded (967 steam generator tubes were plugged prior to the recent Unit 2
'mid-cycle inspection).
The large break LOCA methodology, that is being added to TS 6.9.1. 10 with this proposed amendment (CENPD-132, Supplement 3-P-A, June 1985),
.is the methodology being used to justify Unit,2 operation with greater than 1100 plugged steam generator tubes and is the methodology used to establish the core operating limits for restart from the plant's current mid-cycle outage.
This new methodology will also be used for Units 1 and 3
beginning with their next reloads.
The amendment to Section 6.9. 1.10 adds an analytical method supplement to the analytical methods used to determine the core operating limits.
This analytical method has been previously approved by the staff as defining an acceptable large break model in compliance with Appendix K to 10 CFR Part 50.
The staff's approval letter directs that submittals demonstrating compliance with Appendix K to 10 CFR Part 50, using the Combustion Engineering model after publication of the revised CENPD-132, should reference the model described in the Safety Evaluation enclosed with the NRC letter of July 31, 1986.
Consequently, the licensee proposed to delete those versions of CENPD-132, and the associated NRC approval letter references, presently included in TS Section 6.9. 1. 10 that will be supplanted by Supplement 3-P-A.
The staff discussed with the licensee several of the assumptions as stated in the July 31, 1986, Safety Evaluation of Combustion Engineering's model for emergency core cooling system large break response.
The model is applicable to CE designs.supplied with CE manufactured,zircaloy fuel.
All of the Palo Verde units utilize CE manufactured zircaloy fuel.
Palo Verde Unit 3 has a TS allowance to substitute up to a total of 80 fuel rods clad with zirconium-based alloys for in-reactor performance through fuel cycle 6.
These rods are in bundles in non-limiting positions in the core, and are acceptable for use with this new methodology.
The staff also verified that the following assumptions in the staff's original SE were adhered to:
(1) where the heat transfer coefficients resulting from the HCROSS computer program are greater than those resulting from the FLECHT-based correlation, the FLECHT values are utilized; (2) the limiting break flow discharge coefficient has been determined by an appropriate break spectrum (three guillotine and three slot breaks);
(3) although the homogenous equilibrim break flow model was discussed
.in the topical, the Appendix K Hoody model was used for predicting, break flows; and (4) an axial power shape similar to Shape 8 was utilized.
Since the staff'as reviewed and approved CENPD-132, Supplement 3-P-A, June 1985, this change to TS Section 6.9. 1. 10 is considered to be administrative in nature.
Plant operation will continue to be limited in accordance with values of cycle specific parameters established using NRC-approved methodologies.
il t
The staff has reviewed the proposed changes to TS 6.9. 1. 10 and, based on the above evaluation, finds them acceptable for operation of Palo Verde Nuclear Generating Station Units 1, 2, and 3.
3.0 STATE CONSULTATION
In accordance with the Commission s regulations, the Arizona State official was notified of the proposed issuance of the amendments.
The State official had no comments.
4.0 ENVIRONMENTAL CONSIDERATION
The amendments change a requi,rement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.
The. NRC staff has determined. that the amendments involve no significant increase, in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.
The Commission has previously. issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (59 FR 46069).
Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment
.need be prepared in connection with the issuance of the amendments.
5.0 CONCLUSION
The Commission has concluded, based on the considerations discussed
- above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered, by operation in the proposed
- manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the, issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor:
B. Holian Date:
October 7,
1994
0
~,
~