ML20045C686
| ML20045C686 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 06/16/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20045C682 | List: |
| References | |
| NUDOCS 9306240179 | |
| Download: ML20045C686 (5) | |
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UNITED STATES j
Wfj NUCLEAR REGULATORY COMMISS{ON f'
WASHINGTON. D C. 20666-0001
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.156 TO FACILITY OPERATING LICENSE NO. DPR-20 CONSUMERS POWER COMPANY PALISADES PLANT DOCKET NO. 50-255
1.0 INTRODUCTION
By letter dated January 29, 1993, as supplemented on April 20, 1993, the Consumers Power Company (the licensee) requested an amendment to the Technical Specifications (TS) appended to facility Operating License No. DPR-20 for the Palisades Plant.
The proposed amendment would revise the Palisades TS Table 3.23-2, Radial Peaking Factor Limits, to add limits for those new fuel bundles to be installed during the 1993 Cycle 11 refueling outage.
In addition, the bases for several Specifications (2.1, 2.3, 3.1, 3.12, and 3.23.2) have been updated to reflect the revision of analytical reports for Cycle 11.
The April 20, 1993, submittal provided a correction to the original submittal and did not change the initial proposed no significant hazards consideration determination.
The evaluation for Cycle 11 operation is provided in the Siemens Power Corporation (SPC) report EMF-92-178 entitled, " Palisades Cycle 11:
Disposition and Analysis of Standard Review Plan Chapter 15 Events." This report documents the results of a disposition and analysis of the FSAR Chapter 14 events in support of Palisades Cycle 11 operation with up to 15% steam generator tube plugging.
The events were evaluated in accordance with Chapter 15 of the Standard Review Plan (SRP) and SPC methodology.
The proposed changes for Cycle 11 include (1) the insertion of the third full reload of fuel that uses High Thermal Performance (HTP) grid spacers; (2) increase in assembly and rod radial power peaking limits to accommodate a low radial leakage loading pattern; and (3) the reinsertion of eight Reload N partial shielding assemblies (PSA) and sixteen Reload I hafnium assemblies in low
-powered peripheral locations to reduce vessel fluence.
2.0 EVALUATION The system transients for non-LOCA events were previously analyzed for Cycle 9.
The licensee identified that the Cycle 11 changes (core loading and increase in radial peaking limits) affect only the event minimum departure from nucleate boiling ratio (MDNBR). Therefore, the licensee concluded that the system thermal hydraulic response for the Cycle 9 non-LOCA transient analysis remains valid for Cycle 11.
The large break loss of coolant' accident (LBLOCA) was analyzed previously with radial peaking limits consistent with those for Cycle 11 and it remains bounding (Reference 2).
9306240179 930616 PDR ADOCK 05000255 i
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.The licensee reviewed the Chapter 15 analyses and selected those events that required reanalysis. Their basis for event selection is documented in the Disposition and Analysis of Events report (Reference 1).
Listed below are the SRP Chapter 15 events affecting the nuclear steam supply system that were reanalyzed for the Cycle 11 submittal:
Increase In Heat Removal by the Secondary System 15.1.3 Increase in Steam Flow Decrease in Reactor Coolant System Flow 15.3.1 Loss of Forced Reactor Coolant Flow 15.3.3 Reactor Coolant Pump Rotor Seizure Reactivity and Power Distribution Anomalies 15.4.2 Uncontrolled Control Rod Bank Withdrawal at Power Operation Conditions 15.4.3 Control Rod Misoperation (1) Dropped Control Bank / Rod (2)
Single Control Rod Withdrawal Decreases in Reactor Coolant Inventory 15.6.1 Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve Of the events listed above, two are not bounded by the Cycle 10 analysis.
They include the reactor coolant pump (RCP) rotor seizure and single control rod withdrawal events. The evaluations of these events are discussed below.
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2.1 Reactor Coolant Pumo Rotor Seizure The RCP rotor seizure accident causes the pump to stop, reducing core flow, resulting in a reactor scram on low flow.
With the reduction in flow, the i
primary coolant temperature rises causing the power to rise.
The subsequent temperature and power rise challenge thermal limits; therefore, reanalysis of the MDNBR and maximum linear heat rate (LHR) is required.
4 The licensee calculated the MDNBR for RCP rotor seizure as 1.14 and the peak pellet LHR as 15.9 kW/ft.
The calculated MDNBR is below the ANFP correlation limit of 1.15.
The licensee predicts 0.1% fuel failure due to the violation of the DNBR limit. The licensee indicated that the _ radiological consequences of this amount of fuel failure are a very small fraction of the 10 CFR 100 limits.
-2.2 Sinale Control Rod Withdrawal The rod withdrawal event is initiated by an electrical or mechanical failure in the Rod Control System that causes the inadvertent withdrawal of a single control rod. The movement of a single rod out of sequence causes an insertion of positive reactivity and a local increase in the radial power peaking factor.
The combinations of these factors challenge the DNB margin, therefore, reanalysis of the MDNBR and LHR was performed.
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. The licensee calculated the MDNBR for single control rod withdrawal event to be 1.19 and the peak LHR to be 18.5 kw/ft.
For this event, the MDNBR is greater than the 95/95 DNBR limit for the ANFP correlation and the peak LHR is less than the 21 kW/ft limit for centerline melt.
3.0 CONCLUSION
The licensee has determined that some of the Chapter 15 accident analyses required reanalysis due to Cycle 11, Reload 0.
Of the events that required reanalysis, two were reviewed in this safety evaluation - Single Control Rod Withdrawal (SRP 15.4.3) and RCP Rotor Seizure (SRP 15.3.3).
The staff has reviewed the submittal; in summary, the analyses predicted that the MDNBR will decrease and peak LHR will increase for the single rod withdrawal event in comparison to the Cycle 10 analysis. The predicted LHR for RCP rotor seizure increased over the Cycle 10 analysis.
The predicted MDNBR limit for RCP rotor seizure is below the AFNP correlation limit, and the associated fuel failure is predicted to be an acceptably low value of 0.1% of all fuel pins.
Based on the submittal, the staff has concluded that the specified acceptable fuel design limits for the single rod withdrawal event would be met; namely, the fuel shall not experience centerline melt, i.e., LHR is less than 21 Kw/ft, and the DNBR shall have a minimum allowable limit such that there is a 95% probability with a 95% confidence interval that DNB has not occurred.
Although the predicted MDNBR is less than the 1.15 limit for the RCP rotor seizure accident, the licensee has satisfied the acceptance critera in that the potential radiological consequences are within the limits of 10 CFR 100.
Therefore, the staff finds the proposed changes to the Cycle 11 radial peaking factors limits acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Michigan State Official was notified of the proposed issuance of the amendment.
The Michigan State Official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
l 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 staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any l
effluents that may be released offsite, and that there is no significant l
increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment l
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. involves no significant hazards consideration and there has been no public comment on such finding (58 FR 19476). 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 staff 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 i
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:
S. Brewer, SRXB Dated: are 16.1993 l
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7.0 REFERENCES
1.
Letter from G.B. Slade, Consumers Power, to USNRC, " Palisades Plant Technical Specification Change Request for Cycle 11," dated January 29, 1993.
2.
EMF-91-177, Siemens Nuclear Power Corporation, " Palisades Large Break LOCA/ECCS and Analysis With Increased Radial Peaking and Reduced ECCS Flow," dated October 1991.
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