ML14183A112

Safety Evaluation Supporting Amend 115 to License DPR-23

ML14183A112
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Site:	Robinson
Issue date:	03/07/1988
From:	Office of Nuclear Reactor Regulation
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ML14183A111	List: ML14183A110 ML14183A112
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NUDOCS 8803110008
Download: ML14183A112 (8)

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o UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 115 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 letters dated February 24, 26 and March 1, 1988, the Carolina Power &

Light Company submitted a request for changes to the H. B. Robinson Steam Electric Plant, Unit No. 2, Technical Specifications.

The amendment changes the Technical Specifications to restrict the steady state reactor core power level to less than 1380 megawatts thermal (MWt) when only two safety injection pumps, each capable of automatic initiation from a separate emergency bus, are operable. The Technical Specification change also requires that, prior to exceeding 1380 MWt, NRC review and approval is required.

2.0 DISCUSSION AND EVALUATION There are three safety injection (SI) pumps provided for H. B. Robinson, Unit 2. The current licensing basis for the plant takes credit for two of the three SI pumps to mitigate the design basis loss-of-coolant accident (LOCA). The SI pump A is powered from train A emergency bus E-1, while the pump C is powered from the redundant train B emergency bus E-2.

The third "swing" pump B is normally energized from the E-1 through circuit breaker 22B. However, in the event of a failure of train A, the swing pump also can be energized from the train B emergency bus E-2 (through circuit breaker 298) by utilizing an automatic bus transfer scheme. To prevent simultaneous closure of these circuit breakers (228 and 29B), interlocks are provided to ensure adequate independence of redundant trains.

With the present design configuration, the licensee discovered that a single failure of the diesel generator voltage regulator could result in loss of two of the three SI pumps (i.e., SI pump A and B).

With only one SI pump the facility cannot satisfy the requirements of 10 CFR 50.46 while operating at full power. To eliminate this single failure vulnerability, the licensee has proposed a two SI pump configuration where two SI pumps (A&C) will be required operable under the proposed Technical Specification change, one SI pump on each train. In addition, the licensee will be required to limit the operating power level to 60% of the rated power. At

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2 this reduced power level, the licensee's LOCA analysis has determined that only one SI pump (assuming a single failure of the other SI pump) will be sufficient to mitigate the LOCA. Thus, the modification (Mod. 951) requires removing the auto start capability of SI "swing" pump B and opening of the SI pump B breaker 29C. The interlocks between circuit breakers (22B and 29B) and the auto sequences of two SI pumps (A&C) will remain unchanged. Although not required for the design basis accident, the SI pump B could be manually loaded and powered from either emergency bus.

To determine the allowable power level for the plant with only one SI pump for ECCS performance, after the assumption of a single failure, the licensee evaluated the depressurization events which may need the SI flow to mitigate the consequence during transients. The transients evaluated were: (1) inadvertent operation of a steam generator PORV or safety valve, (2) main steamline break, (3) feedwater line break, (4) inadvertent operation of the ECCS, and (5) steam generator tube rupture. Based on its evaluation of analytical results for the depressurization transients in the FSAR at selected power levels, the licensee found that the impact of operation of SI pumps on the results of the transient is insignificant even for full power operation. Therefore, the licensee concluded, and the staff agreed, that use of one SI pump does not change the conclusions for the current FSAR transient analysis, which assumed two SI pumps available for SI flow.

2.1 Large Break Loss-of-Coolant-Accident Analysis Evaluation The licensee provided the results of a large break LOCA analysis supporting the request for operation up to 1380 MWt (60% power).

The licensee analyzed and evaluated the double ended cold leg guillotine (DECLG) break with a discharge coefficient of 0.4, since this break was identified previously as the limiting case resulting in the highest peak cladding temperature (PCT). The DECLG break analysis was performed with a power peaking factor (F ) of 2.26, 102% of the full power of 2300 MWt, and an assumed lo9s of offsite power at the beginning of the accident. To satisfy the worst single failure criteria, the licensee assumed only one SI pump available in the analysis. The flow changes resulting from the use of one SI pump rather than two have no affect on the limiting break pipe. The analysis was performed by using the modified version of the 1981 Westinghouse ECCS evaluation model with inclusion of the BART methods, which were previously approved by NRC.

The staff has reviewed the large break LOCA analysis and found that (1) the calculated PCT is 2198.50 F which is less than the acceptance criteria of 22000 F, (2) the maximum local metal-water reaction is 7.14 percent which is below the limit of 17 percent, and (3) total core metal-water reaction is less than 0.3 percent which does not exceed the acceptable limit of 1.0 percent. Since approved methods and computer codes were used and the analytical results are within the acceptance criteria of 10 CFR 50.46, this analysis would demonstrate that for a large break LOCA the plant can still satisfy

-3 10 CFR 50.46 at full power, with some adjustment of the peaking factor, with one SI pump (two pumps but assuming a single failure).

The case at 100% power bounds the case at 60% power because of the lower fuel temperature, lower decay heat and lower stored energy in the core at 60% power, and the large break LOCA analysis presented for 2300 MWt is adequate to support plant operation at 1380 MWt (60% power).

2.2 Small Break Loss-of-Coolant-Accident Analysis Evaluation The limiting accident for the revised SI configuration is the small break LOCA. Initially, there had been some indication that the plant might still satisfy 10 CFR 50.46 at or near full power relying on manual actuation of the swing SI pump. However, after further review the licensee subsequently proposed the changes discussed below including a restriction in power to 1380 MWt.

The small break LOCA (SBLOCA) analysis was performed with the approved codes; i.e., (1) NOTRUMP for the calculation of the transient depressurization of RCS, core power, water-steam mixture height and steam flow past the uncovered portion of the core, and (2) LOCTA for the PCT analysis. Three small break LOCA analyses were done assuming 102% of 1380 MWt (60% of full power) and assuming one HHSI pump available for delivery of the SI flow. These analyses were performed for 2.0-inch, 1.5-inch and 1.0-inch equivalent diameter breaks. The 2.0-inch case had the highest PCT of 965.40 F.

An analysis was also performed for a 3.0-inch break, which was previously identified as the limiting SBLOCA case, with full power and only one SI pump available. The results for the 3.0 inch break at full power showed that the PCT is 1772' F, which is within the acceptance criteria of 22000 F. The licensee indicated, and the staff agreed, that the 3.0-inch case at 100% power will bound the 3.0-inch case at 60% power.

The staff has concluded that the small break LOCA analyses are acceptable since the approved method was used, a sufficient break spectrum was analyzed, and the analytical results for all cases of operation at 1380 MWt are within the acceptance criteria of 10 CFR 50.46.

2.3 Technical Specifications Changes The evaluation of the Technical Specifications changes submitted follows:

(1) Technical Specification 3.3.1.1.c The licensee proposed adding a note to this section, which restricts the operating power up to 1380 MWt for the conditions with only two SI pumps operable (each capable of automatic initiation from a separate emergency bus).

This change is

-4 acceptable since the change is supported by the acceptable analysis discussed in this evaluation. In addition, the note states that, prior to exceeding 1380 MWt, NRC review and approval is required. This condition is acceptable to assure adequate emergency core cooling capability to support operation above 1380 MWt.

(2) Technical Specifications 3.3.1.2.b Additional surveillance requirements and corrective action are provided for operation up to 1380 MWt assuming the loss of one of the two SI pumps which is required to be operable. The changes are consistent with the current Technical Specifications required for loss of an SI pump and are acceptable.

(3) Technical Specifications 3.10.2.1, 3.10.2.2, 3.10.2.2.1, 3.10.2.2.2 and Figure 3.10-3.

Notes are added to the related sections regarding reduction of the power peaking factor (F ) from 2.32 to 2.26 for conditions with only two SI pumps operable above 50% of full power. The corresponding value for power levels less than 50% of full power decreases from 4.64 to 4.52. The corresponding Axial Power Distribution Monitoring System value is reduced proportionally from 2.103 to 2.049. These changes are consistent with the assumptions used in the supporting analysis and are acceptable.

3.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION

The Commission has provided standards for determining whether or not a no significant hazards consideration exists (10 CFR 50.92(c)). A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from an accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

The following evaluation in relation to the three standards demonstrates that the proposed amendment does not involve a significant hazards consideration.

1. The proposed amendment would not involve a significant increase in the probability or consequences of any accident previously evaluated.

Prolonged operation at 1380 MWt would not result in significant changes in the flow conditions of the reactor coolant system that could increase the probability of an accident. The H. B. Robinson plant had an extensive history of reduced power operation prior to the steam generator replacement in 1984 and experienced no condition that could increase the probability of an accident.

-5 As indicated above, the changes affect large break and small break LOCA accident sequences and have little or no affect on other accidents and transients. The proposed changes do not increase the probability of either a large or small break LOCA. The SI system is a part of the emergency core cooling systems designed to mitigate the effects of loss of coolant accidents in the event such accidents should occur. The changed configuration does not affect the probability of pipe rupture or any other initiating event leading to a loss of coolant. The power restriction compensates for the elimination of automatic transfer of the swing SI pump and assures that plant operation will satisfy emergency core cooling system requirements with adequate reliability to satisfy the single failure requirements of 10 CFR 50.46 and Appendix K and of 10 CFR Part 50 Appendix A General Design Criterion 35.

With power limited to 1380 MWt, in the event of either a large break or a small break LOCA, the calculated peak clad temperature will be within the limits of 10 CFR 50.46(b), using widely used calculational methods which have been previously approved by the NRC staff. This will assure that the consequences of the only accidents affected by the changes will not significantly increase over those previously analyzed in the Analysis Section of the H.B. Robinson

.Final Safety Analysis Report (UFSAR). In fact, the power restriction should serve to reduce calculated consequences somewhat.

2. The SI system is a part of the emergency core cooling systems designed to mitigate the effects of loss of coolant accidents in the event such accidents should occur. The changed configuration does not affect the probability of pipe rupture or any other initiating event leading to a loss of coolant. Its only affect is on SI system response to previously analyzed accident sequences and as discussed above, with the compensating power restriction, the changes involved in this amendment do not significantly increase the probability or consequences of such previously analyzed accidents.

3. Operation of the facility, in accordance with the proposed amendment, would not involve a significant reduction in a margin of safety.

The analysis of reduced power operation has shown that postulated failures will not produce plant conditions which exceed the safety parameters specified in the Accident Analysis of the UFSAR.

Specifically, there is no significant reduction in safety margin on the reactor core parameters, such as peak fuel clad temperatures, during postulated accidents for the proposed amendment in comparison with those prior to the amendment for full power operation with three operable SI pumps.

Based on the foregoing, the Commission has concluded that the standards of 10 CFR 50.92 are satisfied. Therefore, the Commission has made a final determination that the proposed amendment does not involve a significant hazards consideration.

-6 4.0 FINDING ON EXISTENCE OF EMERGENCY SITUATION The regulations at 10 CFR 50.91(a)(5) provide the necessary requirements for issuing an amendment when the Commission finds that an emergency situation exists and failure to act in a timely way would result in derating or shutdown of a nuclear plant. The Commission expects its

)icensees to: apply for license amendments in a timely fashion; not abuse the emergency provisions by failing to make a timely application for the amendment and thus itself creating the emergency; and provide an explana tion as to why the emergency situation occurred and why it could not have been avoided.

The H. B. Robinson plant has been shutdown since January 29, 1988, when the licensee identified that the SI system did not meet the single failure criterion to support full power operation. This previously unanalyzed condition was identified during the licensee's review of the SI system control logic in response to an NRC request for information related to the emergency electrical distribution to the SI pumps. The NRC request was made on January 14, 1988.

The basic design for emergency electrical distribution to the SI system has not been changed since the plant was licensed to operate. Prior to the review, the licensee had no knowledge that the single failure criterion could not be met for full power operation. Once the problem was identified and the plant placed in cold shutdown, the licensee promptly evaluated and made several modifications to the emergency distribution system control logic which would restore the SI system for full power operation under a number of single failure scenarios. However, following the completion of modifications on February 16, 1988, there remains one scenario for which the licensee has not identified a method for resolution. The licensee also informed the NRC staff that near-term resolution to this remaining single failure scenario is not expected. In order for the plant to restart with only two operable SI pumps, evaluation shows that the power level has to be restricted to no more than 1380 MWt. Consequently, by letters dated February 24, February 26, 1988 and March 1, 1988, the licensee requested that an emergency amendment to place restrictions on power level be processed to allow the plant to restart when only two SI pumps are operable.

Unrelated to the SI system, the licensee has experienced over-speed trips of the emergency diesel generators (EDGs) during fast start tests. The plant will not be restarted until the EDGs have been repaired and their operability verified. The licensee's schedule for plant heatup is on or about March 4, 1988. Therefore, an emergency license amendment is required to avoid delay of startup of the plant.

The staff has reviewed the licensee's explanation of the circumstances justifying consideration of this amendment on an emergency basis. Based on this review, the staff finds that the licensee used its best efforts to apply for the subject amendment in a timely manner and that it had not acted in a manner to create the emergency to take advantage of these procedures.

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5.0 ENVIRONMENTAL CONSIDERATION

This amendment changed a requirement with respect to 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 effluents that may be released off site; and that there is no significant increase in individual or cumulative occupational radiation exposure. 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.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above that:

(1) these amendments will not (a) significantly increase the probability or consequences of accidents previously evaluated, (b) create the possibility of a new or different accident from any previously evaluated, or (c) significantly reduce a margin of safety and, therefore; the amendments do not involve significant hazards considerations; (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; and (3) such activities will be conducted in compliance with the Commission's regula tions and the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.

The staff consulted with the State of South Carolina and the State of South Carolina did not have any comments.

Principal Contributors:

Robert Jones Summer Sun Peter Kang Dated:

March 7, 1988

AMENDMENT NO. 115 TO FACILITY OPERATING LICENSE NO. DPR-23 H. B. ROBINSON, UNIT 2 DISTRIBUTION:

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Wanda Jones E. Butcher (insert name of Principal Contributor of SE ACRS (10)

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