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Notice

Notice of Opportunity To Comment on Model Safety Evaluation on Technical Specification Improvement To Revise the Completion Time From 1 Hour To 24 Hours for Condition B of Technical Specification 3.5.1, “Accumulators,” and Its Associated Bases, Using the Consolidated Line Item Improvement Process

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Start Preamble

AGENCY:

Nuclear Regulatory Commission.

ACTION:

Request for comment.

SUMMARY:

Notice is hereby given that the staff of the Nuclear Regulatory Commission (NRC) has prepared a model safety evaluation (SE) relating to the modification of the completion time from 1 hour to 24 hours for Condition B of Technical Specification (TS) 3.5.1, “Accumulators,” and its associated Bases. The NRC staff has also prepared a model no significant hazards consideration (NSHC) determination relating to this matter. The purpose of these models is to permit the NRC to efficiently process amendments that propose to revise the completion time from 1 hour to 24 hours for Condition B of TS 3.5.1, “Accumulators,” and its associated Bases. Licensees of nuclear power reactors to which the models apply could request amendments confirming the applicability of the SE and NSHC determination to their reactors. The NRC staff is requesting comments on the model SE and model NSHC determination prior to announcing their availability for referencing in license amendment applications.

DATES:

The comment period expires August 14, 2002. Comments received after this date will be considered if it is practical to do so, but the Commission is able to ensure consideration only for comments received on or before this date.

ADDRESSES:

Comments may be submitted either electronically or via U.S. mail.

Submit written comments to: Chief, Rules and Directives Branch, Division of Administrative Services, Office of Administration, Mail Stop: T-6 D59, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

Hand deliver comments to: 11545 Rockville Pike, Rockville, Maryland, between 7:45 a.m. and 4:15 p.m. on Federal workdays.

Copies of comments received may be examined at the NRC's Public Document Room, 11555 Rockville Pike (Room O-1F21), Rockville, MD.

Comments may be submitted by electronic mail to CLIIP@nrc.gov.

Start Further Info

FOR FURTHER INFORMATION CONTACT:

Girija Shukla, Project Manager, Mail Stop: O-7E1, Division of Licensing Project Management, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone 301-415-8439.

End Further Info End Preamble Start Supplemental Information

SUPPLEMENTARY INFORMATION:

Background

Regulatory Issue Summary 2000-06, “Consolidated Line Item Improvement Process for Adopting Standard Technical Specification Changes for Power Reactors,” was issued on March 20, 2000. The consolidated line item improvement process (CLIIP) is intended to improve the efficiency of NRC licensing processes. This is accomplished by processing proposed changes to the standard technical specifications (STS) in a manner that supports subsequent license amendment applications. The CLIIP includes an opportunity for the public to comment on proposed changes to the STS following a preliminary assessment by the NRC staff and finding that the change will likely be offered for adoption by licensees. This notice is soliciting comment on a proposed change to the STS that modifies requirements regarding missed surveillances. The CLIIP directs the NRC staff to evaluate any comments received for a proposed change to the STS and to either reconsider the change or to proceed with announcing the availability of the change for proposed adoption by licensees. Those licensees opting to apply for the subject change to technical specifications are responsible for reviewing the staff's evaluation, referencing the applicable technical justifications, and providing any necessary plant-specific information. Each amendment application made in response to the notice of availability would be processed and noticed in accordance with applicable rules and NRC procedures.

This notice involves the revision of the accumulators completion time from 1 hour to 24 hours in TSs. This proposed change was proposed for incorporation into the STSs by all Owners Groups participants in the Technical Specification Task Force (TSTF) and is designated TSTF-370. TSTF-370 can be viewed on the NRC's Web page at http://www.nrc.gov/​reactors/​operating/​licensing/​techspecs/​.

Applicability

This proposed change to modify TS to revise the accumulators completion time from 1 hour to 24 hours is applicable to all Westinghouse nuclear steam supply system (NSSS) plants regardless of plant vintage and number of loops.

The CLIIP does not prevent licensees from requesting an alternative approach or proposing the changes without the attached model SE and the NSHC. Variations from the approach recommended in this notice may, however, require additional review by the NRC staff and may increase the time and resources needed for the review.

Public Notices

This notice requests comments from interested members of the public within 30 days of the date of publication in the Federal Register. Following the staff's evaluation of comments received as a result of this notice, the staff may reconsider the proposed change or may proceed with announcing the availability of the change in a subsequent notice (perhaps with some changes to the safety evaluation or proposed no significant hazards consideration determination as a result of public comments). If the staff announces the availability of the change, licensees wishing to adopt the change will submit an application in Start Printed Page 46543accordance with applicable rules and other regulatory requirements. The staff will in turn issue for each application a notice of consideration of issuance of amendment to facility operating license(s), a proposed no significant hazards consideration determination, and an opportunity for a hearing. A notice of issuance of an amendment to operating license(s) will also be issued to announce the revision to the completion time for Condition B of TS 3.5.1, “Accumulators,” and its associated Bases for each plant that applies for and receives the requested change.

Proposed Safety Evaluation

U.S. Nuclear Regulatory Commission

Office of Nuclear Reactor Regulation

Consolidated Line Item Improvement, Technical Specification Task Force (TSTF) Change TSTF-370, Risk-Informed Evaluation of an Extension to Accumulator Completion Times for Westinghouse Plants

1.0 INTRODUCTION

The Nuclear Energy Institute (NEI) Technical Specification Task Force (TSTF) has proposed a generic change to the standard technical specifications (STSs) (NUREG-1431) on behalf of the industry. This proposed generic technical specifications (TSs) change, identified by TSTF-370, will revise the completion time (CT) from 1 hour to 24 hours for Condition B of Technical Specification (TS) 3.5.1, “Accumulators,” and its associated Bases. Condition B of TS 3.5.1 currently specifies a CT of one hour to restore a reactor coolant system (RCS) accumulator to operable status when declared inoperable due to any reason except not being within the required boron concentration range.

2.0 BACKGROUND

Topical Report WCAP-15049, “Risk-Informed Evaluation of an Extension to Accumulator Completion Times,” was submitted to the NRC on August 20, 1998, and approved in the NRC letter dated February 19, 1999. The WCAP evaluates the risk associated with extending the accumulator CT from 1 hour to 24 hours for reasons other than boron concentration out of specification.

Wolf Creek was the lead plant for the Westinghouse Owners Group (WOG) program and received plant specific approval for changes to the TSs on April 27, 1999 (License Amendment No. 124). In the NRC letter of February 19, 1999, the staff indicated that it will not repeat its review of the matters described in Topical Report WCAP-15049 when the report appears as a reference in license applications, except to ensure that the material presented applies to the specified plants involved.

The proposed change revises the CT from 1 hour to 24 hours for Condition B of TS 3.5.1, “Accumulators,” and its associated Bases. Condition B of TS 3.5.1 currently specifies a CT of one hour to restore a RCS accumulator to operable status when declared inoperable due to any reason except not being within the required boron concentration range.

3.0 EVALUATION

Deterministic Evaluation

The purpose of the emergency core cooling system (ECCS) accumulators is to supply water to the reactor vessel during the blowdown phase of a loss-of-coolant accident (LOCA). The accumulators are large volume tanks, filled with borated water and pressurized with nitrogen. The cover-pressure is less than that of the reactor coolant system so that following an accident, when the reactor coolant system pressure decreases below tank pressure, the accumulators inject the borated water into the RCS cold legs. The current deterministic safety analysis has not been changed, and thus the limiting condition of operation (LCO), i.e., the lowest functional capability required for safe operation continues to be:

“LCO 3.5.1, [Four] ECCS accumulators shall be operable.

Applicability: Modes 1 and 2, Mode 3 with RCS pressure > [1000] psig.”

Where the bracketed information is nominal, and is subject to substitution of plant specific values.

Under Actions, TSs allow for limited deviations from the LCO. Historically, these Actions and associated CTs have been set using judgement and are not part of the deterministic safety analysis discussed above. Currently, the TS allows for one accumulator to be inoperable for one hour for reasons other than boron concentration not within limits during Modes 1, 2, and in Mode 3 with pressurizer pressure > a plant specific pressure. The WCAP, as well as this TSTF, proposes to increase this CT to 24 hours. The proposed CT of 24 hours is an extension of the current ACTION statement. CTs are by their nature determined by conditions of risk and the impact of the proposed change on risk is reviewed in the following section.

Risk Evaluation

A three-tiered approach, consistent with RG 1.177,[1] was used by the staff to evaluate the risk associated with the proposed accumulator CT, or allowed outage time (AOT), extension from 1 hour to 24 hours. The need for the proposed change was that the current one-hour CT would be insufficient in most cases for licensees to take a reasonable action when an accumulator was found to be inoperable.

Tier 1: Quality of Probabilistic Risk Assessment (PRA) and Risk Impact

Westinghouse used a reasonable approach to assess the risk impact of the proposed accumulator CT extension. The approach is generally consistent with the intent of the applicable NRC RGs 1.174 [2] and 1.177. The quantitative risk measures addressed in the topical report included the change in core damage frequency (CDF) and incremental conditional core damage probability (ICCDP [3] ) for a single CT. The change in large early release frequency (LERF) and incremental conditional large early probability (ICLERP[4] ) for a single CT was qualitatively addressed. Representative calculations were performed to determine the risk impact of the proposed change. Various accumulator success criteria were considered in these calculations to encompass the whole spectrum of Westinghouse plants, e.g., two-, three- and four-loop plants. A reasonable effort was also made to address the differences in other components of risk analysis such as initiating event (IE) frequency and accumulator unavailability among Westinghouse plants.

Westinghouse considered a comprehensive range of IEs in the risk analysis. LOCAs in all sizes—large, medium and small—were included, and reactor vessel failure and interfacing system LOCA were also considered. Modeling of accumulators for mitigation of events other than large, medium and small LOCAs was identified to have insignificant risk impact; therefore, the analysis was performed only on Start Printed Page 46544accumulator injection in response to large, medium and small LOCA events.

The success criteria considered are summarized as follows:

LOCA categoryNo. of loopsSuccess criteria
Large43 accumulators to 3 of 3 intact loops (3/3); 2 accumulators to 2 of 3 intact loops (2/3); no accumulators required (0/3).
32 accumulators to 2 of 2 intact loops (2/2); 1 accumulator to 1 of 2 intact loops (1/2); no accumulators required (0/2).
Medium and Small21 accumulator to 1 of 1 intact loop (1/1); no accumulators required (0/1).
43 accumulators to 3 of 3 intact loops (3/3).
32 accumulators to 2 of 2 intact loops (2/2).
21 accumulator to 1 of 1 intact loop (1/1).

The success criteria considered in this analysis were comprehensive and considered conservative in many cases. For example, many plants indicated the accumulator success criteria for medium and small LOCA events resulted from their role in an alternate success path, in which high pressure injection (HPI) had already failed. Additionally, the staff's review of a number of the original individual plant examinations (IPEs) indicated that no accumulator was needed at all for many medium LOCA sequences and for most of small LOCA sequences.

The fault trees that model accumulator unavailabilities were evaluated. The assumptions made in the fault tree modeling were detailed and were found to be reasonable. For example, the model assumed that the total CT would be used for each corrective maintenance, and this was considered conservative. A comprehensive list of failure mechanisms was considered, and potential common cause failures for check valves and motor-operated valves were also included. Westinghouse used the Multiple Greek Letter technique to determine the common cause failure contributions to the accumulator injection failure.

The component failure rates were taken from the Advanced Light Water Utility Requirements Document.[5] Accumulator unavailabilities due to boron concentration out of limit and due to other reasons were calculated based on a survey of a number of Westinghouse plants. The values for component failure rates and accumulator unavailabilities were within reasonable range. The common cause factors used were also comparable to those used in other PRAs. The accumulator fault trees were quantified using the WesSAGE computer code. The code provided information on the unavailability and cutsets related to the component failures and maintenance activities modeled in the fault trees. A separate hand calculation was used to determine the unavailability due to potential common cause failures. Evaluation of some of the cutsets provided in the topical report did not reveal any unexpected results.

The staff examined the accident sequence identification for each LOCA category. The probability of the sequence leading to core damage involving accumulator failure is summarized for each LOCA category as follows:

Large LOCA: (Large LOCA IE frequency) x (accumulator unavailability).

Medium LOCA: (Medium LOCA IE frequency) x (unavailability of HPI) x (accumulator unavailability).

Small LOCA: (Small LOCA IE frequency) x (unavailability of HPI) x (accumulator unavailability).

The LOCA IE frequencies used for WCAP-15049 are summarized below. Also listed are the LOCA frequencies used in NUREG/CR-4550 [6] (the NUREG-1150 study) for pressurized water reactors (PWRs) and those in the original IPEs.

WCAP-15049NUREG-1150IPE Average (High; Low)
Large LOCA3 x−4/yr5 x 10−4/yr3.3 x 10−4/yr (5 x 10−4/yr; 1 x 10−5/yr).
Medium LOCA8 x 10−4/yr1 x 10−3/yr7.9 x 10−4/yr (2.6 x 10−3/yr; 1 x 10−4/yr).
Small LOCA7 x 10−3/yr1 x 10−3/yr8.9 x 10−3/yr (2.9 x 10−2/yr; 3.7 x 10−4/yr).

Westinghouse indicated that the IE frequencies for WCAP-15049 were based on the plant-specific information contained in the Westinghouse Owners Group (WOG) PSA Comparison Database, which documented the PRA modeling methods and results of the updated PRAs for Westinghouse plants. The mean IE frequencies were used for the risk analysis. These were comparable to the values used for the NUREG-1150 study and the average values in the original IPEs. The staff also found that the IE frequency values in high range among the original IPEs were not much higher than those used for this topical report. The HPI unavailability values used were 7 x 10−3 and 1 x 10−3/yr for medium and small LOCA events, respectively. The staff's examination revealed that the HPI unavailability values were generally comparable to those used in other PRAs, and were generally conservative.

The risk measures calculated to determine the impact on plant risk were based on three different cases. The risk measures considered in each case included the impact on CDF and ICCDP for a single CT, and the impact on LERF and ICLERP for a single CT were qualitatively considered. The three cases considered were:

Design basis case. This case required accumulator injection only for mitigation of large LOCA events (3/3 for 4-loop, 2/2 for 3-loop, and 1/1 for 2-loop).

Case 1. This case credited realistic accumulator success criteria (2/3 for 4-loop, 1/2 for 3-loop, and 0/1 for 2-loop) for large LOCA events and credited the use of accumulators in responding to medium and small LOCA events (3/3, 2/Start Printed Page 465452, and 1/1 for 4-loop, 3-loop, and 2-loop, respectively) following failure of HPI.

Case 2. This case credited more realistic improved accumulator success criteria (no accumulator required) for large LOCA events and credited the use of accumulators in responding to medium and small LOCA events (3/3, 2/2, and 1/1 for 4-loop, 3-loop, and 2-loop, respectively) following failure of HPI.

The results were summarized as follows:

CaseLOCA CDF (/yr) (Current)LOCA CDF (/yr) (Proposed)ΔCDFICCDP
4-loop Design Basis6.93 x 10−79.24 x 10−72.31 x 10−78.20 x 10−7
4-loop Case 16.23 x 10−87.77 x 10−81.54 x 10−85.53 x 10−8
4-loop Case 24.57 x 10−86.09 x 10−81.52 x 10−85.41 x 10−8
3-loop Design Basis4.62 x 10−76.18 x 10−71.56 x 10−78.21 x 10−7
3-loop Case 14.27 x 10−85.31 x 10−81.04 x 10−85.48 x 10−8
3-loop Case 23.05 x 10−84.08 x 10−81.03 x 10−85.42 x 10−8
2-loop Design Basis2.31 x 10−73.09 x 10−77.80 x 10−88.21 x 10−7
2-loop Case 11.52 x 10−82.04 x 10−85.20 x 10−95.42 x 10−8
2-loop Case 21.52 x 10−82.04 x 10−85.20 x 10−95.42 x 10−8

For both realistic cases, the ΔCDFs and ICCDPs were very small for 2-loop, 3-loop, and 4-loop plants, and were much below the numerical guidelines in the RGs 1.174 and 1.177. The staff also noted that the values were considered still bounding in the sense that the risk analysis used a multitude of conservative assumptions and data in the modeling. For many Westinghouse plants, the realistic impact on risk would be much smaller than the values above.

A set of sensitivity cases were also calculated using higher IE frequencies for small and medium LOCAs. The results of the sensitivity calculations did not cause the overall risk impact to increase significantly.

Westinghouse indicated that accumulator success or failure has no direct impact on the containment performance, and that the LERF would therefore increase only in direct proportion to the increased CDF due to accumulator failures. Westinghouse concluded that, since the impact on CDF was small, the impact on LERF would also be small. The staff found the Westinghouse argument to be acceptable; therefore, the impact on LERF and ICLERP for a single CT was very small.

One of the potential benefits of the proposed extended CT was the averted risk associated with avoiding a forced plant shutdown and startup. The risk associated with a forced plant shutdown and ensuing startup due to the inflexibility in current TS could be significant in comparison with the risk increase due to the proposed accumulator CT increase.

Based on the staff's Tier 1 review, the quality of risk analysis used to calculate the risk impact of the proposed accumulator CT extension was reasonable and generally conservative. It was also found that the risk impact of the proposed change was below the staff guidelines in RGs 1.174 and 1.177.

Tier 2 and 3: Configuration Risk Control

Tier 2 of RG 1.177 addresses the need to preclude potentially high risk configurations which could result if certain equipment is taken out-of-service during implementation of the proposed TS change (in this case accumulator CT). If such configurations are identified, the licensee should also identify appropriate measures to avoid them.

The accumulators are always needed to mitigate large size LOCAs. Large LOCAs require accumulators to inject as analyzed under Tier 1 in order to avoid core damage. This means that if a large LOCA occurs without the accumulator function, the core will be damaged independently of whether other systems, such as HPI, function properly or not. However, the probability that a large LOCA occurs in the 24-hour CT is extremely small (in the order of 1E-7 or less). Furthermore, no compensatory or other measures are possible. Due to the negligible risk increase associated with this scenario and the fact that there are no measures to take once a large LOCA occurs, no “high risk” configurations are associated with this scenario.

In general, medium LOCAs do not require accumulators if at least one HPI train is available. This means that if a medium LOCA occurs when minimum accumulator functionality is unavailable and at the same time HPI is unavailable, the core will be damaged. However, the probability that a medium LOCA occurs in the 24-hour CT and at the same time both trains of HPI are unavailable is extremely small (in the order of 1E-8 or less), because it is assumed that the plant is not operating at power with both HPI trains out-of-service. This assumption is based on current STS that limit operation at power with no HPI capability. Therefore, no Tier 2 restrictions beyond those currently in the STS are deemed necessary.

Tier 3 calls for a program to identify “risk significant” configurations beyond those identified in Tier 2 resulting from maintenance or other operational activities and take appropriate compensatory measures to avoid such configurations. Because the accumulator sequence modeling is relatively independent of that for other systems, the Tier 2 analysis by itself is sufficient.

Furthermore, 10 CFR 50.65(a)(4) (Maintenance Rule) requires that licensees assess the risk any time maintenance is being considered on safety-related equipment. This requirement serves the objectives of Tier 3.

In summary, the Tier 2 evaluation did not identify the need for any additional constraints or compensatory actions that, if implemented, would avoid or reduce the probability of a risk-significant configuration. The current TS provisions were found to be sufficient to address the Tier 2 issue. Because the accumulator sequence modeling is relatively independent of that for other systems and the implementation of the Maintenance Rule, the staff concluded that application of Tier 3 to the proposed accumulator CT was not necessary.

The NRC staff finds that the proposed changes will allow safe operation with the changes in CT from 1 hour to 24 hours for Condition B of TS LCO 3.5.1, “Accumulators,” and its associated Bases. The NRC staff also finds that the proposed changes are consistent with the incremental conditional core damage probabilities calculated in WCAP-15049 for the accumulator allowed outage time increase and meet the criterion of 5E-07 in RGs 1.174 and 1.177. The analysis and acceptance provided in this SE, as demonstrated by WCAP-15049, covers all Westinghouse NSSS plants regardless of plant vintage and number of loops. The NRC staff, therefore, concludes that the proposed Start Printed Page 46546TSTF-370, Revision 0 changes are acceptable.

5.0 STATE CONSULTATION

In accordance with the Commission's regulations, the [ ] State official was notified of the proposed issuance of the amendment. The State official had [(1) no comments or (2) the following comments—with subsequent disposition by the staff].

6.0 ENVIRONMENTAL CONSIDERATION

The amendment changes 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 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 ( FR ). 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. se

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.

Proposed No Significant Hazards Consideration Determination

Description of Amendment Request: The proposed amendment would change the technical specifications to revise the completion time (CT) from 1 hour to 24 hours for Condition B of TS 3.5.1, “Accumulators,” and its associated Bases. Condition B of TS 3.5.1 currently specifies a CT of one hour to restore a reactor coolant system (RCS) accumulator to operable status when declared inoperable due to any reason except not being within the required boron concentration range.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below:

Criterion 1—The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated

The basis for the accumulator limiting condition for operation (LCO), as discussed in Bases Section 3.5.1, is to ensure that a sufficient volume of borated water will be immediately forced into the core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators, thereby providing the initial cooling mechanism during large RCS pipe ruptures. As described in Section 9.2 of the WCAP-15049, “Risk-Informed Evaluation of an Extension to Accumulator Completion Times,” evaluation, the proposed change will allow plant operation in a configuration outside the design basis for up to 24 hours, instead of 1 hour, before being required to begin shutdown. The impact of the increase in the accumulator CT on core damage frequency for all the cases evaluated in WCAP-15049 is within the acceptance limit of 1.0E-06/yr for a total plant core damage frequency (CDF) less than 1.0E-03/yr. The incremental conditional core damage probabilities calculated in WCAP-15049 for the accumulator CT increase meet the criterion of 5E-07 in Regulatory Guides (RG) 1.174 and 1.177 for all cases except those that are based on design basis success criteria. As indicated in WCAP-15049, design basis accumulator success criteria are not considered necessary to mitigate large break loss-of-coolant accident (LOCA) events, and were only included in the WCAP-15049 evaluation as a worst case data point. In addition, WCAP-15049 states that the NRC has indicated that an incremental conditional core damage frequency (ICCDP) greater than 5E-07 does not necessarily mean the change is unacceptable.

The proposed technical specification change does not involve any hardware changes nor does it affect the probability of any event initiators. There will be no change to normal plant operating parameters, engineered safety feature (ESF) actuation setpoints, accident mitigation capabilities, accident analysis assumptions or inputs.

Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Criterion 2—The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From Any Previously Evaluated

No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures are introduced as a result of the proposed change. As described in Section 9.1 of the WCAP-15049 evaluation, the plant design will not be changed with this proposed technical specification CT increase. All safety systems still function in the same manner and there is no additional reliance on additional systems or procedures. The proposed accumulator CT increase has a very small impact on core damage frequency. The WCAP-15049 evaluation demonstrates that the small increase in risk due to increasing the accumulator allowed outage time (AOT) is within the acceptance criteria provided in RGs 1.174 and 1.177. No new accidents or transients can be introduced with the requested change and the likelihood of an accident or transient is not impacted.

The malfunction of safety related equipment, assumed to be operable in the accident analyses, would not be caused as a result of the proposed technical specification change. No new failure mode has been created and no new equipment performance burdens are imposed.

Therefore, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Criterion 3—The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety

The proposed change does not involve a significant reduction in a margin of safety. There will be no change to the departure from nucleate boiling ratio (DNBR) correlation limit, the design DNBR limits, or the safety analysis DNBR limits.

The basis for the accumulator LCO, as discussed in Bases Section 3.5.1, is to ensure that a sufficient volume of borated water will be immediately forced into the core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators, thereby providing the initial cooling mechanism during large RCS pipe ruptures. As described in Section 9.2 of the WCAP-15049 evaluation, the proposed change will allow plant operation in a configuration outside the design basis for up to 24 hours, instead of 1 hour, before being required to begin shutdown. The impact of this on plant risk was evaluated and found to be very small. That is, Start Printed Page 46547increasing the time the accumulators will be unavailable to respond to a large LOCA event, assuming accumulators are needed to mitigate the design basis event, has a very small impact on plant risk. Since the frequency of a design basis large LOCA (a large LOCA with loss of offsite power) would be significantly lower than the large LOCA frequency of the WCAP-15049 evaluation, the impact of increasing the accumulator CT from 1 hour to 24 hours on plant risk due to a design basis large LOCA would be significantly less than the plant risk increase presented in the WCAP-15049 evaluation.

Therefore, this change does not involve a significant reduction in a margin of safety.

Based upon the reasoning presented above and the previous discussion of the amendment request, the requested change does not involve a significant hazards consideration.

Start Signature

Dated at Rockville, Maryland, this 9th day of July, 2002.

For the Nuclear Regulatory Commission.

Robert L. Dennig,

Chief, Technical Specifications Section, Operating Reactor Improvements Program, Division of Regulatory Improvement Programs, Office of Nuclear Reactor Regulation.

End Signature End Supplemental Information

Footnotes

1.  RG 1.177, “An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications,” September 1998.

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2.   RG 1.174, “An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,” July 1998.

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3.   ICCDP = [(conditional CDF with the subject equipment out-of-service)—(baseline CDF with nominal expected equipment unavailabilities) x (duration of single CT under consideration)].

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4.   ICLERP = [(conditional LERF with the subject equipment out-of-service)—(baseline LERF with nominal expected equipment unavailabilities) x (duration of single CT under consideration)].

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5.  “Advanced Light Water Utility Requirements Document,” Volume II, ALWR Evolutionary Plant, Chapter 1, Appendix A, PRA Key Assumptions and Ground Rules, Rev. 5, Issued December 1992.

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6.  NUREG/CR-4550, “Analysis of Core Damage Frequency: Internal Events Methodology,” Vol. 1, Rev. 1, January 1990.

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[FR Doc. 02-17649 Filed 7-12-02; 8:45 am]

BILLING CODE 7590-01-P