ML20072G020
| ML20072G020 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 08/12/1994 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20072G017 | List: |
| References | |
| NUDOCS 9408240170 | |
| Download: ML20072G020 (4) | |
Text
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UNITED STATES 5
NUCLEAR REGULATORY COMMISSION k..s
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4 WASHINGTON. D.C. 20555-0001 l
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 65 TO FACILITY OPERATING LICENSE NO. NPF-51.
ARIZONA PUBLIC SERVICE COMPANY. ET AL.
PALO VERDE NUCLEAR GENERATING STATION. UNIT NO. 2 DOCKET N0. STN 50-529
1.0 INTRODUCTION
By letter dated July 1, 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, Unit 2 (Appendix A to Facility Operating License No. NPF-51). 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 Water and Power, and Southern California Public Power Authority. The proposed changes would modify TS Figure 3.2-1, " REACTOR COOLANT Cold-leg vs CORE POWER LEVEL," of TS 3/4.2.6, " REACTOR COOLANT Cold-leg TEMPERATURE," to include the cold-leg temperature between 552 *F and 562 *F at core power levels between 90 percent and 100 percent within the AREA 0F ACCEPTABLE OPERATION. This TS change would permit reactor operation at full power with a lower reactor coolant temperature to minimize potential steam generator tube degradation due to primary water stress corrosion cracking (PWSCC).
By letter dated August 11, 1994, the licensee supplemented its application.
The additional information contained in this letter was clarifying in nature, was within the scope of the initial notice, and did not affect the NRC staff's 4
proposed no significant hazards determination.
2.0 BACKGROUND
The cold-leg temperature reduction at power levels above 90 percent was previously granted for Units 1 and 3 by letter dated June 7, 1994.
The change was not requested for Unit 2 due to continuing analysis arising from steam
. generator tube plugging in that unit. Currently, Figure 3.2-1 of the Palo Verde Unit 2 TS specifies a minimum cold-leg temperature of 562 *F for core.
power levels between 90 percent and 100 percent.
This TS requirement is supported by the safety analyses documented in the Updated Final Safety Analysis Report (UFSAR). Operation of Palo Verde Uni _t 2 is currently restricted by an administrative 1y imposed 10 *F reduction in the reactor coolant cold-leg temperature, by maintaining reactor power at approximately 88 percent of rated core power to' satisfy the current TS requirements. The licensee requested a change to the TS to allow Unit 2 to operate at full power g0824017o940012 p
ADOCM 05000529 PDR
, 3 with a 10 *F deduction in the reactor coolant cold-and hot-leg temperatures.
The hot-leg temperature reduction is needed to minimize the steam generator tube degradation (e.g., PWSCC).
3.0 EVALUATION In support of its proposed changes to the TS, the licensee submitted the results of an evaluation for all transients and accident analyses documented in the UFSAR.with the reduced cold-leg temperature.
For events that were no longer bounded by the analyses presented in UFSAR, new analyses were performed to ensure that the acceptance criteria are still met for each event.
For transients involving an increase in heat removed by the secondary system or a decrease in reactor coolant flow rate, the UFSAR analysis covers cold-leg temperatures of 550 *F and above; therefore, no new analysis is needed.
Regarding transients involving a decrease in heat removal by the secondary system, the loss of condenser vacuum event is affected by a reduction in cold-leg temperature and was reanalyzed assuming a cold-leg temperature of 550 *F.
The results of the analysis were acceptable.
The consequences of a main feedwater line break are more limiting with higher values of cold-leg temperature; therefore, this event was not reanalyzed.
Transients involving reactivity and power distribution anomalies were reviewed for power levels above 90 percent. Acceptable results are assured because adequate thermal margin remains available with the reduced cold-leg temperature.
The transients involving an increase in reactor coolant system inventory are determined to_ be independent of the value of cold-leg temperature. Similarly, transients involving a decrease in reactor coolant system inventory are not affected because either a higher cold-leg temperature is more limiting or the results of analysis are independent of cold-leg temperature.
The licensee also performed a reanalysis for a postulated main steam line break with induced multiple steam generator tube ruptures, assuming cold-leg temperatures of 550 *F and 540 *F.
The results demonstrate that the Palo 1
Verde administrative limits for primary iodine activity will ensure that the dose consequences remain acceptable.
Additionally, cycle-specific loss-of-coolant accident (LOCA) evaluations were performed for Unit 2, Cycle 5.
These evaluations included a reduction in cold-leg temperature of 10 *F.
The resultant peak clad temperature remains less than the current analysis of record limit of 2091 *F and the licensing i
basis limit of 2200 *F.
Due to the higher number of plugged steam generator tubes in Unit 2, the licensee performed additional analyses. to examine LOCA results taking into account a greater tube plugging limit. This analysis examined operation of Unit 2 during Cycle 5 at-100 percent power with up to 1100 steam generator tubes plugged (with no restriction on the number of plugged tubes in each steam generator).
The analysis determined that the results of the reference t..
..j
. 2 analysis documented in the PVNGS Unit 2 Cycle 5 Reload Analysis Report conservatively apply at a peak linear heat generation rate of 13.2 KW/ft, a reduction from 13.5 KW/ft that has subsequently been incorporated in the Unit 2 Core Operating Limits Report (COLR). The analysis explicitly evaluated the impact of the increase in the number of plugged steam generator tubes on the refill /reflood hydraulic transient of the limiting large break LOCA and the consequential impact on the fuel cladding temperature and maximum local cladding oxidation of the hot rod.
Regarding small-break LOCA scenarios, the licensee concluded that the peak cladding temperature of the limiting small-break LOCA continues to be hundreds of degrees less than that of the limiting large-break LOCA. Additionally, the licensee stated that the post-LOCA long term cooling analysis, which currently extends to 1000 plugged tubes for each steam generator, conservatively bounds the 1100 total plugged tube analysis.
The staff noted that currently Unit 2 steam generators (SGs) have a total of approximately 967 tubes plugged (226 in SG 21; 741 in SG 22). Therefore, the LOCA analysis performed with an assumption of 1100 tubes plugged provides a margin of approximately 133 tubes.
Additionally, each of the last two outages for eddy current inspection has resulted in tube plugging rates whereby the tube plugging margin may be exceeded during the next scheduled outage (approximately 248 tubes were plugged during U2R4; and 409 during U2M5).
The licensee stated that a contingency plan is being evaluated (including, potentially, further reducing the peak linear heat generation limit) in case the upcoming outage results in exceeding the steam generator tube plugging limit assumed in the analysis. Additionally, the staff notes that the non-LOCA analyses were reanalyzed with the assumption of 1000 tubes plugged per SG. The licensee committed to inform the staff and summarize corrective actions if either the non-LOCA or LOCA tube-plugging limits are exceeded.
Further, the licensee clarified several items in its letter of August 11, 1994. The licensee's submittal stated that core protection calculator (CPC) coefficients exhibit little sensitivity to a reduction in cold-leg temperature. The licensee supported their conclusion of no negative effect on the design basis event (loss of external load) with a discussion of temperature filter coefficients that provide for lead / lag inputs in transients. Regarding control element assembly drop events, the staff questioned why reactor coolant density changes would not have an adverse affect on drop times.
The licensee stated that the safety analysis drop time is assumed at 4 s-nds, and surveillance data continues to be in the range of 2.5 seconds, maint.1ning adequate margin.
Similar to the calculation necessary for Unit 1, the licensee took credit for negative moderator temperature coefficient (MTC) (even though the TS allow a positive MTC for beginning of cycle operation).
The staff verified that the value of negative MTC is appropriate for this time in cycle length, and notes that core operating limits supervisory system database and/or addressable constants may need to be reverified by the licensee prior to operation with positive MTC values following the next refueling outage.
Finally, the staff questioned why the pressurizer spray events were not dependent on the value of T-cold.
The licensee stated that the events are based on the dynamic response of the CPCs and that analysis shows that, with a 550 *F inlet temperature, the CPCs respond appropriately.
In summary, based on the results of the Unit 2 cycle-specific evaluations and i
the evaluations discussed above, a 10 'F reduction in cold-leg temperature is acceptable for Unit 2.
The staff has reviewed the licensee's submittal and finds that the proposed TS change is supported by appropriate analysis and is acceptable.
4.0 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.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement 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 amendment involves 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 amendment involves no significant hazards consideration, and there has been no public comment on such finding (59 FR 35767). Accordingly, the amendment meets 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 amendment.
6.0 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 Contributors:
B. Holian C. Liang Date: August 12, 1994 4