ML20154F383

Recommends Early NRR Mgt Review of Encl Safety Evaluation Re Plant 25% License

ML20154F383
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	09/14/1988
From:	Varga S Office of Nuclear Reactor Regulation
To:	Crutchfield D Office of Nuclear Reactor Regulation
Shared Package
ML20154F388	List: ML20154F383 ML20154F392 ML20154F395 ML20154F403
References
NUDOCS 8809190373
Download: ML20154F383 (9)

Text

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f SEP 1 '4 M NOTE TO: Dennis M. Crutchfield, Acting Associate Director for Projects FROM: Steven A. Varga, Director Division of Reactor Projects I/II

SUBJECT:

SHOREHAM SER FOR 25% LICENSE Attached is a copy of the subject SER. I recomend its early review by NRR management.

This SER incorporates the Emergency Preparedness Branch's interpretation of FEMA's "reasonable assurance" finding (see side bar). However, yesterday afternoon, 0GC infomed us that the E00 (based on a late Friday afternoon meeting) may be looking for a more comprehensive application of the FEMA finding than provided in the attached version. Because this issue most likely will dominate the Shoreham discussion at the 9/15 ED0 briefing, our approach on this matter should be reviewed by the Executive Team prior to that briefing.

We will continue work, unless directed otherwise, toward issuance of the attached version of the 25% SER as soon as possible. Point of contact is Stewart Brown on 21444.

S Steven A. Varga, Director Division of Reactor Projects I/II

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%; .' . . . . , SEP 14198 NOTE TO: Dennis M. Crutchfield, Acting Associate Director for Projects FROM: Steven A. Varga, Director Division of Reactor Projects I/II StIBJECT: SHOREHAM SER FOR 25% LICENSE Attached is a copy of the subject SER. I recomend its 9arly review by NRR management.

This SER incorporates the Emergency Preparedness Branch's interpretation of FEMA's "reasonable assurance" finding (see side bar). However, yesterday afternoon OGC informed us that the ED0 (based on a ' ate Friday afternoon meeting) may be looking for a more comprehensive application of the FEMA finding than provided in the attached version. Bec3use this issue most likely will dominate the Shoreham discussion at the 9/15 EDO briefing, our approach on this matter should be reviewed by the Executive Team prior to that briefing.

We will continue work, unless directed otherwise, toward issuance of the attached version of the 25% SER as soon as possible. Point of contact is Stewart Brown on 21444 o

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UNITED STAMS NUCLEAR REGULATORY COMMISSION WASHINGTON, 0.0.20665 e

THE NRC STAFF'S TECHNICAL REVIEW OF A REQUEST FROM LONG ISLAND LIGHTING COMPANY FOR AUTHORIZATION TO OPERATE THE SHOREHAM NUCLEAR POWER STATION AT A POWER LEVEL UP TO TWENTY-FIVE PERCENT OF FULL-RATED POWER INTRODUCTION On April 14, 1987, the Long Island Lighting Company (LILCO) submitted to the Comission a request for authorization to increase power to 25 percent of rated power at the Shoreham Nuclear Power Station (the Request). In the Request.

LILCO claims that 10 CFR 50.47 provides the regulatory basis for the authorization for operation beyond 5 percent power, despite the existence of unresolved emergency planning (EP) contentions. In the Request, LILCO seeks to demonstrate that it can meet all three of the conditions set forth in 10 CFR 50.47(c)(1) with the restriction on power level to 26 percent. Specifically, LILCO contends that the implementation of its emergency plari by its local emergency response organization and local governments on a best-efforts basis, coupled with the 25 percent power limitation, constitutes an adequate comper _' ting measure for the interim period when the contested EP issues are still being litigated in regard to the full-power license.

In terms of the power limitation as an "interim compensatirg action," LILC0 claims that the risk and consequences of accidents at 25 pecent power operation are so greatly reduced that the remainino unresolved EP issues become insignificant. The staff's technical review is an attempt to assess the validity of LILCO's claim about this reduction of risks and consequences from the analysis that was submitted with the Request. The staff's evaluation does not examine the unresolved EP' issues and whether the safety merits associated with the 25 percent power restriction would constitute adequate compensating measures. Instead, the emphnis of this technical review is on comparisons between operation at 25 percent power and at full power and the effects of the power reduction on various aspects of postulated accidents.

On September 9, 1988, the .ederal Emergency Management Agency (FEMA) provided its finding on LILCO's offsite emergency response plan for Shoreham. FEMA stated that the full participation exercise conducted on June 7-9, 1988 demnnstrated adequate overall preparednes;, on the part of Local Emergency Response Organization personnel. Therefore, based on the evaluation of the plan and the exercise, FEMA reached a finding of reasonable assurance that the health and safety of the puHic living in the vicinity of the plant can be protected. FEMA's plan review and exercise evaluation were based on the assumptions that in an actual radiological emergency, State and incal officials that have declined to participate in emergency planning will (1) exere.ise their best efforts to protect the health and safety of the public, (2) cooperate with the utility and follow the utility plan, and (3) have the resources sufficient to. implement those portions.of the. utility plan where State and. local response is necessary.

'.s SCOPE OF STAFF'S REVIEW The staff's review addrc3ses the following three categories of issues:

1) Systems and Procedures for Accident Mitigation

2) Accident Evaluation

3) Safety of Prolonged Operation at Twenty-Five Percent Power (1) Systems and Procedures for Accident Mitigation Except for the questions related to EP, all safety issues have been satisfacto-rily resolved for full-power operation. In its analysis to demonstrate that there are reduced risk and accident consequences when operating at 25 percent power compared with operating at full-power LILCO cites several physical and procedural improvements made at the Shoreham Nuclear Power Station (SNPS) since the issuance of its 5 percent power license in July 1985. The staff has reviewed the acceptability of these hardware and p ocedural changes for the credit taken in the accident analysis in support of the Request. The Safety Evaluation prepared by the staff is provided as Enclosure 1.

(2) Accident Evaluation The design basis accidents (DBAs) for full-power operation were addressed '

in Section 15 of the SNPS Final Safety Analysis Report, and the consequences of these accidents would not result in the need for offsite evacuation.

Therefore, only those accidents that are beyond the OAA need to be evaluated.

To support its request, LILCO presented a probabilistic risk analysis (PRA) to show that at 25 percent power (a) the probabilities of core-melt accidents are reduced; (b) the offsite radiological consequences of accidents are reduced; and (c) the timing for key events in the accident progression, e.g., core slump, reactor vessel failure, and releases to the environment, is significantly increased. This consideration is beneficial in two important aspects: first, the available time enhances the opportunity for corrective actiens (e.g.,

correct diagnostics, restoration of core cooling, restoration of ac power) to arrest the accident progression, and secondly, the increased duration between the onset of an accident and releases of radioactive materials 10 the environ-ment will significantly increase the time available for emergen:y responses.

The staff's review of LILCO's PRA-based portion of the request is provided as Enclosure 2.

(3) Safety of Proinnged Operation at Twenty-Five Percent Powe; Prolonged off-nomal operation at 25 percent power may cause instability or other undesirable effects on certain safety-related system!. The staff performed an evaluation concerning the reliability of those systems and equipment for which performance is identified to be power-lev 11 dependent.

The Safety Evaluation on this issue is provided as Enclosure 3. j

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SUMMARY

OF RESULTS The following is a sumary of the significant results of the staff's evaluation as presented in Enclosures 1, 2, and 3.

(1) Systems and Procedures for Accident Mitigation The staff finds the following improvements in equipment and procedures to be acceptable for the credits taken in the risk assessment:

(1) The main condenser as the viable heat sink following a turbine trip.

The majority of anticipated transient initiators for boiling water reactors (BWRs) result from or lead to a turbine trip. With the 25 percent power limitation, availability of the main condenser as the only necessary hwat sink is an important mitigating factor for anticipated transient without scram (ATWS) accidents.

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(2) The standby liquid control system at SNPS, which is designed to ensure an equivalent boron injection capability that is 200 percent of the ATdS

, rule requirement of 10 CFR 50.62.

l (3) Compliance with the ATWS rule for the alternate rod injection and 1 recirculation pump trip capabilities to mitigate ATWS accidents.

(41 The design and installation of the "corium ring," which is intended to channel the molten core debris (corium) directly into the suppression pool for. quenching, even though this hardware modification is not explicitly modelled in the risk assessment.

I (5) The additional AC power supplies that are beyond those installed to meet the requirements of Criterion 17 of Appendix A to 10 CFR Part 50. These additional ac power supplies incluJe the gas turbine, four mobile -

diesel engines, and the Colt diesel-engine-powered gererators that could mitigate or avert station blackout accidents.

(6) The procedure to use the diesel fire pump as a viable cooling source.

l (7) The availability of operator options to gain greater control in accident mitigation actions (e.g., throttling of the low-pressure emergency core ,

cooling system and the condensate flow during ATWS events, the capability to switch the high-pressure coolant injection suction to either the suppression pool or the condensate storage tank, and the enhancement of the automatic depressurization system initiation logic).

5 (2) Accident Evaluation The staff's review of LILCO's PRA-based accident analysis for 25 percent power operation concentrates on comparisons with 100 percent power operation. These comparisons were to determine the validity of LILCO's claim that 25 percent power operation involves significant improvements in tenns of vulnerability to l

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i core damage accidents, additional time to respond to accidents, and reductions of offsite consequences of the postulated accidents. These comparisons were based on the use of the same calculation tools and assumptions and have an advantage over calculations for absolute values because the effects of inherent modelling uncertainties tend to be minimized. Furthennore, the comparisons of the timing of events during an accident and of offsite consequences are deterministic in nature, given an assumed core damage accident progression sequence or assumed release characteristics. The uncertainties associated with the probabilities of the paths of the accident development are not relevant for these deterministic comparisons.

(a) Yulnerability to Core Damage Accidents The LILCO 75 percent power PRA calculated a core-melt frequency reduction of approximately a factor of two Le to 25 percent operation, in conjunction with improvements in plant design and emergency operating procedures. The calculated reduction was about a factor of three for station blackout and ATWS sequences.

For a number of specific plant vulnerabilities to core melt, the staff's I evaluation found that the 25 percent power restriction, in conjunction with improvements in plant design and emergency operating procedures, represents improvements that are in general agreement with LILCO's claim. The staff's evaluation supports LILCO's claim that the overall core melt frequency is reduced at 25 percent power compared with 100 percent power. However, the staff did not verify the absolute magnitude of this reduction.

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(b) Offsite Consequences of Accidents The staff's evaluation found LILCO's claim that offsite radiological ,

consequences are reduced at 25 percent power vs. 100 percent power operation to '

be italid. This is due to two factors: first, a reduction of an approximate factor of four in the fission product inventory available for release in any postulated accident at 25 percent power; second. a significant increase in the time to release because of the reduced heatup rate at 25 percent power, i.e.,

the reduced decay heat lovel.

l The staff's evaluation is in agreement with LILCO's analysis that 'ndicates that there is considerable reduction in the probability of exceeding a given dose at an offsite location, even without evacuation. This is particularly the case for larger doses. The distances over which injury-threatening doses (i.e., 200 rem) would occur are reduced by a factor of abou'. three compared with 100 percent power operation. The staff has performed dose-distance calculations for both 25 percent and 100 percent power using the same code and assumptions. For the sequences representing the bulk of the core-melt accidents at Shoreham the calculated probability of exceeding 200 rem falls off rapidly to small values at distances of about one mile from the SNPS site at 25 percent power, versus about three miles at 100 percent nower.  ;

For the less probable, rapidly evolving accidents, representing about three percent of the core-melt frequency in the Request, the calculated probability of exceeding 200 rem falls off rapidly to small values about two miles from the site at 25 percent power compared with 10 miles at 100 percent power.

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Another important consideration of the accident consequence calculation is the significant additional time available to avoid injury-threatening doses.

Dose-versus-time calculations performed for a rapidly evolving accident sequence show that a 200-rem whole-body dose would not be reached at a two mile radius within six hours after accident initiation, in comparison to about one hour in the case of 100 percent power operation.

(c) Timir] of Accident Progression The staff's evaluation agrees with '.(LCO's claim that operation at 25 percent  ;

power would result in considerable delay in accident progression when compared '

with similar accidents occurring at 100 percent power operation. The staff found that significant delays would occur in all postulated accident sequences -

and at every stage of accident development. The major mitigating factor is the 25 percent power limitation and the associated reduction of decay heat that is i the driving force in accident progression, i

For a large group of accidents, characterized by delayed challenges to the containment integrity, releases to the environment at 25 percent power would not occur until well over 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after accident initiation. These accidents include those initiated by a loss of offsite power, the majority of loss-of-coolant accidents, and those transient-initiated sequences for which the reactor is successfully shutdown but core cooling is inadequate. These accidents contribute over 80 percent of the total core-melt frequency. Under more optimistic assumptions regarding reactor vessel failure, core-concrete interactions, .and containment performance, the time of releases to the environment for these sequences at 25 percent power would be on the order of a i

, day or more. For 100 percent power operation, these accidents generally lead '

to radioactive releases in the order of several hours (the majority in the four-to-seven-hour range).

The most rapidly developing accidents are those characterized by early containment failure or containment bypass releases. The dominant accident in this category is the seismically incuced accident that breaches the reactor coolant boundary as well as the containment. The staff estimates that the time from the onset of the accident to the time when radioactive releases to the environment occur is about one hour. The corresponding time estimated for 100 j percent power is about ten minutes.

i The remaining category of accidents is dominated by those involving transients with failure to scram the reactor. The staff estimates that radioactive releases to the environment for this category of accidents for the 25 percent power case is about seven to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> from the onset of the transient. For the case of 100 percent power operation, the higher decay heat represents an earlier challenge to the containment integrity, and releases are estimated to occur in about two and one half hours. l l

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(3) Safety-Related Systems Evaluation The staff agrees with LILCO's evaluation that all safety-related equipment is intended to be operated over the entire power range. However, the staff is concerned if reduced power operation would cause accelerated wear or early fatigue damage to certain safety-related equipment from low-flow-induced vibration or instability. The staff found three systems--the reactor recirculation, the main steam, and the feedwater systems--to be power dependent  ;

and to operate at a reduced flow. In particular, the feedwater check valves are most vulnerable and would serve as a good indicator of any potential equipment deterioration. The staff determines that these check valves should be subject to a more frequent inservice testing inspection schedule; that is, these valves should be inspected during each refueling outage and more frequently than every two years.

CONCLUSION The following are the major findings of the staff's evaluation for 25 percent power operation at SNpS:

(1) There are no new unresolved safety questions associated with 25 percent power operation that have not been analyzed during the full-power licensing review process.

(2) The improvements in equipment and procedures are acceptable for the credits taken in the accident analysis as presented in the Request.

(3) The staff is in general agreement with LILCO's claim that operation at 25 percent power would reduce core-melt frequency. ,

(4) There are significant delays in the time of progression for all the i postulated accident sequences when compared with those during 100 percent power operation:

(a) For accidents contributing to about 80 percent of the 25 percent power core-melt frequency, a long time is recuired for core melt and vessel failure, and radioactive releases would not occur in less than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

(b) The most rapidly developing accidents are those associated with a

rapid loss of coolant and failure of all injection systems. A seismic event that breaches the reactor coolant system and the containment is a representative sequence of this type. The probability of these accidents occurring is small accounting for

! about three percent of core-melt frequency in the LILCO PRA for 25 percent power. The onset of releases are delayed from about 12

! minutes in the case of 100 percent power operation to an hour at 25 percent power. The bulk of radiological releases would occur later:

one hour at full power and three hours at 25 percent power, a

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. 7 (c) The remaining category of accidents in terms of adequate timing to ,

take mitigating action is dominated by ATWS sequences. Radioactive i releases to the environment for these accidents are estimated to '

occur in about seven to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This compares with an estimate of ,

about two and one half hours for similar accidents during 100 percent power operation.

(5) The equilibrium radionuclide inventories are reduced by about a factor of four when compared with 100 percent power operation; offsite radiological consequences can also be expected to be reduced by the same factor.

(6) The distances from the SNPS site within which injury-threatening radioactive doses could occur without evacuation have been significantly reduced. In staff calculations performed with the same code and assumptions, the staff fourd that these distances have been reduced to one or two miles for 25 percent power operation. While the vulnerable areas have been reduced, time available for evacuation from the reduced areas has been significantly increased in the case of 25 percent power operation at SNPS over 100 psrcent power.

(7) Certain plant components could be adversely effected by prolonged operation at a reduced power level due to reduced system flow condition.

These components should be subject to a more frequent inservice testing progran than required for 100 percent power operation.

RIFEREbC[

Letter to Victor Stello (NRC) frow Grant Peterson (FEMA), dated September 9, 1988.

Principal Contributor: R. Lo 1

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