ML14183A312
| ML14183A312 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 04/19/1995 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML14183A311 | List: |
| References | |
| NUDOCS 9504240302 | |
| Download: ML14183A312 (5) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 163 TO FACILITY OPERATING LICENSE NO. DPR-23 CAROLINA POWER & LIGHT COMPANY H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261
1.0 INTRODUCTION
By letter dated April 13, 1995, as supplemented April 18, 1995, Carolina Power
& Light Company (CP&L or the licensee) submitted a request for changes to the H. B. Robinson Steam Electric Plant, Unit No. 2 (HBR), Technical Specifications (TS).
The requested changes would revise TS Section 4.4.3.f, g, and h to allow the post accident heat removal system surveillance test interval to be changed from a 12-month interval to a refueling outage interval.
The present 12-month testing interval for TS 4.4.3 f, g and h was defined in the original HBR license issued in July 1970 when fuel cycle duration was also 12 months. On August 1981, the operating license was amended to include emergency core cooling system (ECCS) leakage minimization surveillance on a refueling interval basis; however, that amendment did not revise the 12-month interval for the post accident recirculation heat removal system surveillance test in TS 4.4.3 that specifies a limit of 2 gallons per hour leakage to the environment. The post accident recirculation heat removal system is comprised of portions of the residual heat removal system (RHR) subject to recirculation flow following an accident. This test requires visual inspection of valve packing, pump seals, and other components for leakage when the system is pressurized to 350 psig.
2.0 EVALUATION The original test interval was based on engineering judgment and was designed to assure system integrity and functionality. The test requires that the system be pressurized to 350 psig and a visual inspection for leakage be conducted. Historically, the test was performed during power operation by pressurizing the residual heat removal system (RHR) from the chemical and volume control system letdown through a pressure limiting valve.
The licensee has proposed to perform the test every refueling outage and that it should be conducted in the equivalent of the standard technical specification (STS) modes 3 or 4. The STS defines hot standby (mode 3) as a reactivity less.than 0.99, 0 percent rated thermal power, and Tavg greater than or equal to 350 degrees F. The STS definition for hot shutdown (mode 4) is the same as for hot standby, except that Tavg is less than 350 degrees F and greater than 200 degrees F. The licensee does not utilize the 9504240302 950419 PDR ADOCK 05000261 P
2 conventional "mode" designation in the HBR TS. For HBR, "Hot Shutdown" requires the reactor to be subcritical and T
> 200 OF.
Of the modes defined in the HBR TS, "Hot Shutdown" is the closest o conventional modes 3 and 4.
The test when performed in the hot shutdown mode does not require that the RHR system be taken out of service. On the other hand, when the test is performed at power, the RHR system needs to be isolated and, thus, needs to be taken out of service. Therefore, by performing the test in the hot shutdown mode, a slight risk benefit is expected relative to performing it at power during the test period. In addition, the individual components, i.e. RHR system valves and pump seals, are tested on a quarterly basis. This is done when the RHR pumps are tested and the system is pressurized to 150 psig. The system at that time is checked for alignment and affords the opportunity to the licensee's staff to identify leaks in the packing and seals. This is not equal to the level of inspection nor the pressure called for in TS 4.4.3.
However, it provides an indication of the condition of the components and some opportunity for leak detection. In the NRC staff's judgment, the increased time interval for the leakage test will not significantly change the risk balance because the system is not normally operated during power operation and the system does not experience significant wear and tear during the surveillance interval. Therefore, we conclude that the proposed TS change for the post accident recirculation heat removal system leak testing does not increase the risk when performed at refueling outages rather than at a 12-month interval.
Thus, we find the proposed TS change acceptable.
3.0 EMERGENCY CIRCUMSTANCES NRC regulations (10 CFR 50.91(a)(5)) require that whenever an emergency situation exists, a licensee must explain why this emergency situation occurred and why it could not avoid this situation, and the NRC will assess the licensee's reasons for failing to file an application sufficiently in advance of the event.
An emergency situation exists when the NRC's failure to act in a timely way would result in derating or shutdown of a nuclear plant, or in prevention of either resumption of operation or of increase in power output up to the plant's licensed power level. In such cases, the NRC may issue a license amendment involving no significant hazards consideration without prior notice and opportunity for a hearing or for public comment. Also, in such cases, the regulations require that the NRC be particularly sensitive to environmental considerations. The discussion of why this proposed change meets the conditions necessary for emergency is provided below.
The refueling outage was originally scheduled for mid-April 1995, but was rescheduled during the third quarter of 1994, to commence on April 29, 1995.
On April 11, 1995, while at full power, the licensee attempted to perform the test of the post accident recirculation heat removal system to meet the TS required test schedule date of April 19, 1995. Unanticipated and unacceptable reactor coolant system (RCS) leakage exceeding 10 gallons per minute through the RHR system into the refueling water storage tank caused the licensee to suspend the surveillance. The last time that the testing was completed during power operations was in 1993, and that test was conducted successfully.
Therefore, the problem currently experienced with performing this test during
3 power operation was unanticipated. Furthermore, performance of this surveillance test by means other than connecting the RHR System to the RCS is impractical based on flow and dose exposure considerations. Upon approval of this TS change, the TS amendment will be implemented immediately, and the test will be conducted during the upcoming refueling outage, utilizing the RHR system in the shutdown cooling mode.
The NRC staff concludes that an emergency situation exists in that failure to act in these circumstances could be reasonably expected to result in an unnecessary shutdown of HBR. Further, the NRC staff finds that the licensee acted in a timely manner after discovering the situation and has not abused the emergency provisions of 10 CFR 50.91(a)(5).
4.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION
DETERMINATION The Commission's regulations in 10 CFR 50.92 state that the Commission may make a final determination that a license amendment involves no significant hazards consideration if operation of the facility in accordance with the amendment would not:
(1) Involve a significant increase in the probability or consequences an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.
The licensee proposed that the requested TS changes did not involve a significant hazards consideration, stating as follows:
This change does not involve a significant hazards consideration for the following reasons.
- 1.
The requested change does not involve a significant increase in the probability or consequences of an accident previously evaluated. The requested amendment will change the interval for the Residual Heat Removal portion of the Post accident Recirculation Heat Removal System leakage test from a "12-month interval" to "refueling."
Since operation of the ECCS in the recirculation mode of operation is not a precursor to an accident evaluated in the safety analysis report, the probability of occurrence of any accident evaluated in the safety analysis report is unchanged. The dose consequences to the control room operators analyzed in Updated Final Safety Analysis Report (UFSAR) Section 6.4 include a dose component from total ECCS leakage during the recirculation phase. The refueling surveillance interval for the ECCS leakage minimization program has already been reviewed by the NRC, and has been included in the current licensing basis as Operating License (OL) Condition 3.G(2). This OL Condition specifies that an integrated leak test for each system be conducted at a frequency not to exceed refueling cycle intervals.
4 Because the total leakage allowed by TS 4.4.3 is maintained and the lengthening of the surveillance interval is effectively insignificant, this change does not constitute an increase in the consequences of an accident previously analyzed.
- 2.
The requested change does not create the possibility of a new kind of accident from any accident previously evaluated. The change in test frequency does not [affect] the ability of the ECCS leakage minimization program to perform its intended function. No new accident scenarios are introduced by performing the required test while in shutdown conditions. None of the analyzed accident scenarios or assumptions are changed by the extension of this surveillance interval.
Therefore, the possibility or probability of occurrence of any new accident from any accident previously evaluated is unchanged.
- 3.
The requested change does not involve a significant reduction in the margin of safety. The margin of safety, as defined in TS Section 4.4.3 of two gallons per hour, is not reduced by this change since this margin is applied to all post accident recirculation systems. The UFSAR accident analyses do not include a specific contribution to the off-site dose consequences from post accident recirculation leakage. However, the dose consequences to the control room operators analyzed in UFSAR Section 6.4, which was performed in response to Three Mile Island (TMI) Action Item III.D.3.4 and provided to the NRC in a letter dated May 21, 1990, and the NRC's off-site dose consequences in the Safety Evaluation Report (SER) for the license amendment to uprate reactor power from 2200 Megawatts thermal (MWt) to 2300 MWt, both included a dose component from ECCS leakage during the recirculation phase. In the control room dose calculation, a value of four gallons per hour, which is two times the TS assumed two gallons per hour leakage requirement was used for conservatism as discussed in a letter to the NRC dated September 5, 1990.
Increasing the length of the surveillance interval to refueling has no effect on system leakage since the system is not normally operated during power operation and the system does not experience significant wear and tear during the surveillance interval.
The NRC staff has reviewed the licensee's analysis and, based on this review, and for the reasons stated therein, has determined that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff has determined that the amendment request involves no significant hazards consideration.
5.0 STATE CONSULTATION
In accordance with the Commission's regulations, the State of South Carolina official was notified of the proposed issuance of the amendment. The State official had no comments.
5
6.0 ENVIRONMENTAL CONSIDERATION
This amendment changes the surveillance requirements. 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 made a final no significant hazards determination with respect to this amendment.
Accordingly, this.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 this amendment.
7.0 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: L. Lois B. Mozafari Date: April 19, 1995