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Notice

Notice of Opportunity To Comment on Model Safety Evaluation on Technical Specification Improvement To Modify Requirements Regarding LCO 3.10.1, Inservice Leak and Hydrostatic Testing Operation Using the Consolidated Line Item Improvement Process

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Information about this document as published in the Federal Register.

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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 shutdown testing requirements in technical specifications (TS) for Boiling Water Reactors (BWR). 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 modify LCO 3.10.1 that would allow control rod scram time testing to be performed concurrently with inservice leak and hydrostatic testing. Licensees of nuclear power reactors to which the models apply could then request amendments, confirming the applicability of the SE and NSHC determination to their reactors. The NRC staff is requesting comment on the model SE and model NSHC determination prior to announcing their availability for referencing in license amendment applications.

DATES:

The comment period expires September 20, 2006. 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, Maryland. Comments may be submitted by electronic mail to NRCREP@nrc.gov.

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FOR FURTHER INFORMATION CONTACT:

Tim Kobetz, Mail Stop: O-12H2, Division of Inspections and Regional Support, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone 301-415-1932.

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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 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 a proposed change to the STS after a preliminary assessment by the NRC staff and a finding that the change will likely be offered for adoption by licensees. This notice solicits comment on a proposal to modify LCO 3.10.1 that would allow control rod scram time testing to be performed concurrently with inservice leak and hydrostatic testing. 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 announce the availability of the change for adoption by licensees.

This notice involves the modification of LCO 3.10.1 that would allow control rod scram time testing to be performed concurrently with inservice leak and hydrostatic testing. This change was proposed for incorporation into the standard technical specifications by the owners groups participants in the Technical Specification Task Force (TSTF) and is designated TSTF-484. TSTF-484 can be viewed on the NRC's Web page utilizing the Agencywide Documents Access and Management System (ADAMS). ADAMS accession numbers are ML052930102 (TSTF-484 Submittal), ML060970568 (NRC Request for Additional Information, RAI), and ML061560523 (TSTF Response to NRC RAIs). Start Printed Page 48562

Applicability

Licensees opting to apply for this TS change 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 will be processed and noticed in accordance with applicable rules and NRC procedures.

Public Notices

This notice requests comments from interested members of the public within 30 days of the date of publication in the Federal Register. After evaluating the comments received as a result of this notice, the staff will either reconsider the proposed change or announce the availability of the change in a subsequent notice (perhaps with some changes to the safety evaluation or the 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 must submit an application in accordance with applicable rules and other regulatory requirements. For each application the staff will publish a notice of consideration of issuance of amendment to facility operating licenses, a proposed no significant hazards consideration determination, and a notice of opportunity for a hearing. The staff will also publish a notice of issuance of an amendment to an operating license to announce the modification of TS 3.10.1, Inservice Leak and Hydrostatic Testing, for each plant that 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-484, Revision 0, Use of TS 3.10.1 for Scram Time Testing Activities

1.0 Introduction

By application dated [Date], [Name of Licensee] (the licensee) requested changes to the Technical Specifications (TS) for the [Name of Facility].

The proposed changes would revise LCO 3.10.1, and the associated Bases, to expand its scope to include provisions for temperature excursions greater than [200]°F as a consequence of inservice leak and hydrostatic testing, and as a consequence of scram time testing initiated in conjunction with an inservice leak or hydrostatic test, while considering operational conditions to be in Mode 4.

2.0 Regulatory Evaluation

2.1 Inservice Leak and Hydrostatic Testing

The Reactor Coolant System (RCS) serves as a pressure boundary and also serves to provide a flow path for the circulation of coolant past the fuel. In order to maintain RCS integrity, Section XI of the American Society of Mechanical Engineers (ASME) Pressure Vessel Code requires periodic hydrostatic and leakage testing. Hydrostatic tests are required to be performed once every 10 years and Leakage tests are required to be performed each refueling outage. Appendix G to 10 CFR Part 50 states that pressure tests and leak tests of the reactor vessel that are required by Section XI of the American Society of Mechanical Engineers (ASME) Pressure Vessel Code must be completed before the core is critical.

NUREG-1433, General Electric Plants, BWR/4, Revision 3, Standard Technical Specifications (STS) and NUREG-1434, General Electric Plants, BWR/6, Revision 3, STS both currently contain LCO 3.10.1, Inservice Leak and Hydrostatic Testing Operation. LCO 3.10.1 was created to allow for hydrostatic and leakage testing to be conducted while in Mode 4 with average reactor coolant temperature greater than [200]°F provided certain secondary containment LCOs are met.

TSTF-484, Revision 0, Use of TS 3.10.1 for Scram Time Testing Activities, modifies LCO 3.10.1 to allow a licensee to implement LCO 3.10.1 while hydrostatic and leakage testing is being conducted should average reactor coolant temperature exceed [200]°F during testing. This modification does not alter current requirements for hydrostatic and leakage testing as required by Appendix G to 10 CFR part 50.

2.2 Control Rod Scram Time Testing

Control Rods function to control reactor power level and to provide adequate excess negative reactivity to shut down the reactor from any normal operating or accident condition at any time during core life. The control rods are scrammed by using hydraulic pressure exerted by the Control Rod Drive (CRD) system. Criterion 10 of Appendix A to 10 CFR part 50 states that the reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel limits are not exceed during any condition of normal operation, including the effects of anticipated operational occurrences. The scram reactivity used in design basis accidents (DBA) and transient analyses is based on an assumed control rod scram time.

NUREG-1433, General Electric Plants, BWR/4, Revision 3, Standard Technical Specifications (STS) and NUREG-1434, General Electric Plants, BWR/6, Revision 3, STS both currently contain surveillance requirements (SR) to conduct scram time testing when certain conditions are met in order to ensure that Criterion 10 of Appendix A to 10 CFR part 50 is satisfied. SR 3.1.4.1 requires scram time testing to be conducted following a shutdown greater than 120 days while SR 3.1.4.4 requires scram time testing to be conducted following work on the CRD system or following fuel movement within the affected core cell. Both SR must be performed at reactor pressure greater than or equal to [800] psig and prior to initially exceeding 40% rated thermal power (RTP).

TSTF-484, Revision 0, Use of TS 3.10.1 for Scram Time Testing Activities, would modify LCO 3.10.1 to allow SR 3.1.4.1 and SR 3.1.4.4 to be conducted in Mode 4 with average reactor coolant temperature greater than [200]°F. Scram time testing would be performed in accordance with LCO 3.10.4, Single Control Rod Withdrawal—Cold Shutdown. This modification to LCO 3.10.1 does not alter the means of compliance with Criterion 10 of Appendix A to 10 CFR part 50.

3.0 Technical Evaluation

The existing provisions of LCO 3.10.1 allow for hydrostatic and leakage testing to be conducted while in Mode 4 with average reactor coolant temperature greater than [200]°F, while imposing Mode 3 secondary containment requirements. Under the existing provision, LCO 3.10.1 would have to be implemented prior to hydrostatic and leakage testing. As a result, if LCO 3.10.1 was not implemented prior to hydrostatic and leakage testing, hydrostatic and leakage testing would have to be terminated if average reactor coolant temperature exceeded [200]°F during the conduct of the hydrostatic and leakage test. TSTF-484, Revision 0, Use of TS 3.10.1 for Scram Time Testing Activities, modifies LCO 3.10.1 to allow a licensee to implement LCO 3.10.1 while hydrostatic and leakage testing is being conducted should average reactor coolant temperature exceed [200]°F during testing. The modification will allow completion of testing without the potential for interrupting the test in order to reduce reactor vessel pressure, cool the RCS, and restart the test below Start Printed Page 48563[200]°F. Since the current LCO 3.10.1 allows testing to be conducted while in Mode 4 with average reactor coolant temperature greater than [200]°F, the proposed change does not introduce any new operational conditions beyond those currently allowed.

Surveillance Requirements (SR) 3.1.4.1 and SR 3.1.4.4 require that control rod scram time be tested at reactor pressure greater than or equal to [800] psig and before exceeding 40% rated thermal power (RTP). Performance of control rod scram time testing is typically scheduled concurrent with inservice leak or hydrostatic testing while the reactor coolant system (RCS) is pressurized. Because of the number of control rods that must be tested, it is possible for the inservice leak or hydrostatic test to be completed prior to completing the scram time test. Under existing provisions, if scram time testing can not be completed during the LCO 3.10.1 inservice leak or hydrostatic test, scram time testing must be suspended. Additionally, if LCO 3.10.1 is not implemented and average reactor coolant temperature exceeds [200]°F while performing the scram time test, scram time testing must also be suspended. In both situations, scram time testing is resumed during startup prior to exceeding 40% RTP. TSTF-484, Revision 0, Use of TS 3.10.1 for Scram Time Testing Activities, modifies LCO 3.10.1 to allow a licensee to complete scram time testing initiated during inservice leak or hydrostatic testing. As stated earlier, since the current LCO 3.10.1 allows testing to be conducted while in Mode 4 with average reactor coolant temperature greater than [200]°F, the proposed change does not introduce any new operational conditions beyond those currently allowed. Completion of scram time testing prior to reactor criticality and power operations results in a more conservative operating philosophy with attendant potential safety benefits.

It is acceptable to perform other testing concurrent with the inservice leak or hydrostatic test provided that this testing can be performed safely and does not interfere with the leak or hydrostatic test. However, it is not permissible to remain in TS 3.10.1 solely to complete such testing following the completion of inservice leak or hydrostatic testing and scram time testing.

Since the tests are performed with the reactor pressure vessel (RPV) nearly water solid, at low decay heat values, and near Mode 4 conditions, the stored energy in the reactor core will be very low. Small leaks from the RCS would be detected by inspections before a significant loss of inventory occurred. In addition, two low pressure emergency core cooling systems (ECCS) injection/spray subsystems are required to be operable in Mode 4 by TS 3.5.2, ECCS-Shutdown. In the event of a large RCS leak, the RPV would rapidly depressurize and allow operation of the low pressure ECCS. The capability of the low pressure ECCS would be adequate to maintain the fuel covered under the low decay heat conditions during these tests. Also, LCO 3.10.1 requires that secondary containment and standby gas treatment system be operable and capable of handling any airborne radioactivity or steam leaks that may occur during performance of testing.

The protection provided by the normally required Mode 4 applicable LCOs, in addition to the secondary containment requirements required to be met by LCO 3.10.1, minimizes potential consequences in the event of any postulated abnormal event during testing. In addition, the requested modification to LCO 3.10.1 does not create any new modes of operation or operating conditions that are not currently allowed.

4.0 State Consultation

In accordance with the Commission's regulations, the [Name of State] State official was notified of the proposed issuance of the amendment. The State official had [no] comments. [If comments were provided, they should be addressed here].

5.0 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. A significant hazards consideration is attached and is available for public comment. 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 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 amendments will not be inimical to the common defense and security or to the health and safety of the public.

7.0 References

1. NUREG-1433, “General Electric Plants, BWR/4, Revision 3, Standard Technical Specifications (STS)”, August 31, 2003.

2. NUREG-1434, General Electric Plants, BWR/6, Revision 3, Standard Technical Specifications (STS)”, August 31, 2003.

3. Request for Additional Information (RAI) Regarding TSTF-484, April, 7, 2006, ADAMS accession number ML060970568.

4. Response to NRC RAIs Regarding TSTF-484, June 5, 2006, ADAMS accession number ML061560523.

5. TSTF-484 Revision 0, “Use of TS 3.10.1 for Scram Times Testing Activities”, May 5, 2005, ADAMS accession number ML052930102.

Model No Significant Hazards Determination

Description of Amendment Request: The proposed changes would revise LCO 3.10.1, and the associated Bases, to expand its scope to include provisions for temperature excursions greater than [200]°F as a consequence of inservice leak and hydrostatic testing, and as a consequence of scram time testing initiated in conjunction with an inservice leak or hydrostatic test, while considering operational conditions to be in Mode 4.

Basis for No Significant Hazards 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.

Technical Specifications currently allow for operation at greater than [200]°F while imposing MODE 4 requirements in addition to the secondary containment requirements required to be met. Extending the activities that can apply this allowance will not adversely impact the probability or consequences of an accident previously evaluated. Therefore, the proposed 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 accident previously evaluated.

Technical Specifications currently allow for operation at greater than Start Printed Page 48564[200]°F while imposing MODE 4 requirements in addition to the secondary containment requirements required to be met. No new operational conditions beyond those currently allowed by LCO 3.10.1 are introduced. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements or eliminate any existing requirements. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice. Therefore, the proposed 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 a margin of safety.

Technical Specifications currently allow for operation at greater than [200]°F while imposing MODE 4 requirements in addition to the secondary containment requirements required to be met. Extending the activities that can apply this allowance will not adversely impact any margin of safety. Allowing completion of inspections and testing and supporting completion of scram time testing initiated in conjunction with an inservice leak or hydrostatic test prior to power operation results in enhanced safe operations by eliminating unnecessary maneuvers to control reactor temperature and pressure. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, the NRC concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

Principal Contributor: Aron Lewin.

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Dated at Rockville, Maryland this 15th day of August 2006.

For the Nuclear Regulatory Commission.

Timothy Kobetz,

Branch Chief, Technical Specifications Branch, Division of Inspections and Regional Support, Office of Nuclear Reactor Regulation.

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[FR Doc. E6-13715 Filed 8-18-06; 8:45 am]

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