ML25247A259

Final Traveler SE of TSTF-601 (ML25247A259)

ML25247A259
Person / Time
Site:	Technical Specifications Task Force
Issue date:	09/08/2025
From:	NRC/NRR/DSS
To:	Technical Specifications Task Force
Shared Package
ML25247A257	List: ML25247A258 ML25247A259
References
EPID L?2024?PMP?0007, TSTF 601, Rev 1
Download: ML25247A259 (10)

Text

FINAL SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION TECHNICAL SPECIFICATIONS TASK FORCE TRAVELER TSTF-601, REVISION 1 EXTEND SHIELD BUILDING COMPLETION TIME AFTER REFUELING USING THE CONSOLIDATED LINE-ITEM IMPROVEMENT PROCESS (EPID: L2024PMP0007)

1.0 INTRODUCTION

By letter dated March 6, 2025 (Agencywide Documents Access and Management System Accession No. ML25065A247), the Technical Specifications Task Force (TSTF) submitted Traveler TSTF-601, Revision 1, Extend Shield Building Completion Time After Refueling, to the U.S. Nuclear Regulatory Commission (NRC). Traveler TSTF-601 proposed to change the shield building standard technical specifications (STS) for Westinghouse and Combustion Engineering designed pressurized-water reactors (PWRs). Upon approval, this change would be incorporated into future revisions of NUREG-1431 and NUREG-1432, and this traveler would be available to licensees for adoption through the consolidated line-item improvement process (CLIIP).1 The proposed change would revise the shield building STS to extend the Completion Time for when the shield building is inoperable while in Modes 4 or 3 following a refueling outage and prior to criticality.

1.1 Shield Building Discussion Some Westinghouse and Combustion Engineering plant designs refer to the shield building using plant-specific terminology, such as the enclosure building and the secondary containment.

For simplicity, throughout this safety evaluation (SE), the STS term shield building is used.

The STS shield building is a concrete structure that surrounds the steel containment vessel.

Between the containment vessel and the shield building inner wall is an annular space that collects containment leakage that may occur following an accident. The shield building in part ensures that the release of radioactive material from the containment atmosphere is restricted to the leakage paths and associated leakage rates assumed in the accident analyses. The annular space also allows for periodic inspection of the outer surface of the containment.

1 *NUREG-1431, Standard Technical Specifications, Westinghouse Plants, Volume 1, Specifications, and Volume 2, Bases, Revision 5, September 2021 (ML21259A155 and ML21259A159, respectively).

• NUREG-1432, Standard Technical Specifications, Combustion Engineering Plants, Volume 1, Specifications, and Volume 2, Bases, Revision 5, September 2021 (ML21258A421 and ML21258A424, respectively).

1.2 Description of the Current Requirements and the Proposed Change Current STS shield building operability requirements NUREG1431, Specification 3.6.8, Shield Building (Dual and Ice Condenser), requires the shield building to be operable in Modes 1, 2, 3, and 4.

NUREG1432, Specification 3.6.11, "Shield Building (Dual)," requires the shield building to be operable in Modes 1, 2, 3, and 4.

In both these specifications, if the shield building is inoperable, it must be restored to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The STS Bases for these specifications explain that 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is a reasonable Completion Time considering the limited leakage design of containment and the low probability of a design-basis accident occurring during this time period.

During a refueling outage, sections of the shield building boundary are removed to allow access to the containment equipment hatch. At the end of the outage after the equipment hatch is installed, restoring the shield building boundary is among the last activities performed prior to entering Mode 4, which places restoration of the shield building on or near the outage critical path.

Proposed change to STS shield building operability requirements Following a refueling outage and prior to criticality, the decay heat generation and the radionuclide inventory are significantly reduced compared to full power operation. As a result, a longer Completion Time to restore the shield building to operable status in Modes 4 and 3 after a refueling and prior to criticality is proposed. TSTF-601, Section 2, Detailed Description, describes the proposed STS change to address an inoperable shield building under these conditions. In particular, the STS shield building specifications would be revised to add a new Action A, as shown below, that applies when the shield building is inoperable while in Mode 3 or 4 following refueling and prior to criticality (Mode 2) with a Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to restore the shield building to operable status:

A. ----------NOTE------------

Only applicable if MODE 2 has not been entered following refueling.

Shield building inoperable in MODE 3 or 4 following refueling.

A.1 Restore shield building to OPERABLE status.

72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> The existing Condition A, which states, Shield Building inoperable, is renamed Condition B and modified to state, Shield building inoperable for reasons other than Condition A. The renamed Condition B Required Action and Completion Time are unaffected. Existing Action B is renamed Action C but otherwise unchanged.

2.0 REGULATORY EVALUATION

As described in the Commissions Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors (58 FR 39132, dated July 22, 1993), [t]he new STS should include greater emphasis on human factors principles in order to add clarity and understanding to the text of the STS, and provide improvements to the Bases Section of the Technical Specifications which provides the purpose for each requirement in the specification.

The improved vendor-specific STS were developed and issued by the NRC in September 1992.

The Summary Section of the Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors states, in part:

Implementation of the Policy Statement through implementation of the improved STS is expected to produce an improvement in the safety of nuclear power plants through the use of more operator-oriented Technical Specifications, improved Technical Specification Bases, reduced action statement induced plant transients, and more efficient use of NRC and industry resources.

Section IV, The Commission Policy, of the Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors states, in part:

The purpose of Technical Specifications is to impose those conditions or limitations upon reactor operation necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety by identifying those features that are of controlling importance to safety and establishing on them certain conditions of operation which cannot be changed without prior Commission approval.

[T]he Commission will also entertain requests to adopt portions of the improved STS [(e.g., TSTF-601)], even if the licensee does not adopt all STS improvements. The Commission encourages all licensees who submit Technical Specification related submittals based on this Policy Statement to emphasize human factors principles.

In accordance with this Policy Statement, improved STS have been developed and will be maintained for each NSSS [nuclear steam supply system] owners group. The Commission encourages licensees to use the improved STS as the basis for plant-specific Technical Specifications. [I]t is the Commission intent that the wording and Bases of the improved STS be used to the extent practicable.

The Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors provides the following description of the scope and the purpose of the STS Bases:

Each LCO [limiting condition for operation], Action, and Surveillance Requirement should have supporting Bases. The Bases should at a minimum address the following questions and cite references to appropriate licensing documentation (e.g., Updated Final Safety Analysis Report (FSAR), Topical Report) to support the Bases.

1. What is the justification for the Technical Specification, i.e., which Policy Statement criterion requires it to be in the Technical Specifications?

2. What are the Bases for each LCO, i.e., why was it determined to be the lowest functional capability or performance level for the system or component in question necessary for safe operation of the facility and, what are the reasons for the Applicability of the LCO?

3. What are the Bases for each Action, i.e., why should this remedial action be taken if the associated LCO cannot be met; how does this Action relate to other Actions associated with the LCO; and what justifies continued operation of the system or component at the reduced state from the state specified in the LCO for the allowed time period?

4. What are the Bases for each Safety Limit?

5. What are the Bases for each Surveillance Requirement and Surveillance Frequency; i.e., what specific functional requirement is the surveillance designed to verify? Why is this surveillance necessary at the specified frequency to assure that the system or component function is maintained, that facility operation will be within the Safety Limits, and that the LCO will be met?

Note: In answering these questions the Bases for each number (e.g., Allowable Value, Response Time, Completion Time, Surveillance Frequency), state, condition, and definition (e.g., operability) should be clearly specified. As an example, a number might be based on engineering judgment, past experience, or PSA [probabilistic safety assessment] insights; but this should be clearly stated.

The regulations in Title 10 of the Code of Federal Regulations (10 CFR) paragraph 50.36(a)(1) require that:

Each applicant for a license authorizing operation of a utilization facility shall include in his application proposed technical specifications in accordance with the requirements of this section. A summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the technical specifications.

The regulations in 10 CFR 50.36(b) require that:

Each license authorizing operation of a utilization facility will include technical specifications. The technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to [10 CFR] 50.34 [Contents of applications; technical information]. The Commission may include such additional technical specifications as the Commission finds appropriate.

The categories of items required to be in the TS are listed in 10 CFR 50.36(c). The regulation at 10 CFR 50.36(c)(2) requires that technical specifications include LCOs. Per 10 CFR 50.36(c)(2)(i), LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility. The regulation also requires that when an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TS until the condition can be met.

The NRC staffs guidance for the review of TS is in NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light -Water Reactor]

Edition (SRP), Chapter 16.0, Technical Specifications, Revision 3, dated March 2010 (ML100351425). As described therein, as part of the regulatory standardization effort, the NRC staff has prepared STS for each of the LWR nuclear designs.

3.0 TECHNICAL EVALUATION

The regulatory framework the NRC staff used to determine the acceptability of the proposed change consists of the requirements and guidance listed in Section 2 of this SE. The NRC staffs review determined whether the proposed change to the STS met the standards for technical specifications (TS) categories in 10 CFR 50.36(c)(2), as well as conformed to the Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors. In addition, the NRC staff reviewed the proposed STS changes for technical clarity and consistency with the existing STS requirements for customary terminology and formatting.

3.1 Evaluation of New Condition A As described in Section 1.2 of this SE, TSTF601 proposed a new Condition A. New Condition A addresses situations where the shield building is inoperable while in Mode 4 or Mode 3 following refueling and before criticality (Mode 2).

TSTF601 states that this change is reasonable given the differences in the plant conditions under the proposed Action compared to the plant conditions at full power. The NRC staff assessed the differences in plant conditions described in TSTF601 such as energy in the core and radionuclide inventory. In addition, the NRC staff assessed the containment and entering the shield building specifications Applicability with the LCO not met.

3.1.1 Energy in the Core TSTF601 assessed the energy in the reactor core that would be present under new Condition A, in part, as follows:

In order to use the proposed 72-hour Completion Time, the reactor will have been shut down for refueling, which takes at least 14 days. ANSI/ANS5.1, Decay Heat Power in Light Water Reactors, indicates that after a 2-week shutdown, the decay heat from fission products would be approximately one-quarter of one percent of the full power heat generation. In addition, during the preceding refueling outage approximately one-third of the core will have been replaced with new, unirradiated fuel. This will further reduce the heat being produced by the core. As a result, the energy in the core under the proposed conditions would be less than that assumed for the [10 CFR 50] Appendix K LOCA analysis by approximately three orders of magnitude.

The NRC staff reviewed the TSTF601 assessment of energy in the reactor core present under new Condition A. The NRC staff concludes that nuclear energy in the reactor core present under new Condition A is much lower when compared to atpower conditions, given the reduction in decay heat and replacement of roughly one-third of the core.

Therefore, based on the above, the NRC staff finds the proposed new Condition A is acceptable in part because of the lower nuclear energy in the reactor.

3.1.2 Accident Dose Consequence Analysis TSTF601 assesses the radionuclide inventory in the reactor that would be present under new Condition A, in part, as follows:

Approximately one-third of the core will be new, unirradiated fuel resulting in a significant reduction of the radioactive material available for release from the core in accident conditions. Also, during the refueling outage, shorter-lived radioisotopes in the irradiated fuel have been reduced due to decay. For example, Iodine-131 (a significant contributor to dose) has a half-life of 8 days and will have been reduced in activity by almost a factor of 4 [four] during a hypothetical 14-day shutdown. These factors reduce the radioactive material available to be released into the shield building, also supporting a longer Completion Time for the proposed conditions.

The NRC staff reviewed the TSTF601 assessment of radionuclides in the reactor core present under new Condition A. As stated in Regulatory Position 3 in NRC Regulatory Guide (RG) 1.183, Alternate Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000 (ML003716792), the core inventory used in design-basis accident loss of coolant accident analysis requires the licensee to assume that the fission products in the reactor core are based on maximum full power operation of the core, and a period of irradiation of sufficient duration to allow all the activity of the dose-significant radionuclides to reach equilibrium or maximum values. The NRC staff concludes that radionuclide inventory available for release in the reactor core present under new Condition A would be significantly lower when compared to atpower maximum hypothetical accident conditions, given the reduction in radionuclide activity due to decay and replacement of roughly one-third of the core.

The transport of radionuclides during the new Condition A would be reduced with the significantly lower energy in the core under the proposed condition. With the Appendix K LOCA analysis under the new Condition A showing that the energy in the core is approximately three orders of magnitude lower than the energy at full power, all pathways to the control room and the environment would have a reduced release. Transport of the source term would be reduced due to the lower reliance on engineered safety features to remove decay heat, and lower decay heat reducing motive force causing primary containment leakage would also result in a lower accident dose consequence.

Multiple plant systems are also required to be operable in Modes 4 and 3 that would significantly lower any radioactive release in the event of an accident while in the proposed Action A, including the primary containment, the containment isolation valves, and the shield building ventilation system. The control room operators would also be protected by the control room emergency filtration system.

Therefore, based on the above, the NRC staff finds the extension of the Completion Time from 24 to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> proposed in the new Condition A is acceptable with respect to a radiological consequences analysis due to (1) the reduction in radioactive material available for release, (2) the operability requirements for other plant systems, and (3) the significantly reduced transport of radionuclides from all release paths to the control room and environment.

3.1.3 Containment TSTF601 points to the current NUREG1431 and NUREG1432 Action Bases for an inoperable shield building. The NUREGs state, in part, that the Completion Time is reasonable considering the limited leakage design of the containment.

The containment structure and its penetrations establish the leakage limiting boundary for the containment function. Maintaining the containment operable when required in Modes 1 through 4 limits the leakage of fission product radioactivity from the containment to the environment during a design-basis accident.

The STS provide containment surveillance requirements to assure that the necessary quality of the containment and its penetrations are maintained, and that containment operation will be within safety limits (e.g., leakage). Leakage testing is performed in accordance with the Containment Leakage Rate Testing Program (a specification in the administrative controls portion of the STS). This STS program addresses compliance with 10 CFR 50, Appendix J. In addition, when the proposed new Condition A is applicable, the containment is required to be operable.

Therefore, based on the above, the NRC staff finds the proposed new Condition A is acceptable, in part, because the STS contain specifications designed to ensure operability of containment prior to entering Mode 4.

3.1.4 Entering the Applicability with the LCO Not Met The proposed change would permit a licensee to enter Modes 4 and 3 following a refueling outage and prior to criticality before completion of the surveillance requirements needed to establish the operability of the shield building.

NUREG1431 and NUREG1432 Surveillance Requirement 3.0.4 established the requirement that all applicable Surveillances must be met before entry into a Mode or other specified condition in the LCOs Applicability. This specification ensures that system and component operability requirements are met before entry into Modes or other specified conditions in the Applicability for which these systems and components ensure safe operation of the unit.

However, a provision in the NUREGs addresses the situation when an LCO is not met due to Surveillances not having been met. Under this situation, entry into a Mode or other specified condition in the Applicability shall only be made in accordance with LCO 3.0.4.

TSTF601 Section 3.3, Entering the Applicability with the LCO Not Met, explains that, under the proposed change, licensees would apply LCO 3.0.4.b to enter Mode 4 and Mode 3 with the shield building inoperable due to a Surveillance (i.e., shield building annulus negative pressure limit) not met. LCO 3.0.4.b permits entry into the Applicability after performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering the Mode or other specified condition in the Applicability, and establishment of risk management actions, if appropriate. The LCO 3.0.4.b risk assessment must consider all inoperable TS equipment, such as the containment, containment isolation valves, the shield building ventilation system, and the control room emergency filtration system. The risk assessment must also consider other conditions such as severe weather and likelihood of weather-generated missiles. The risk assessments will be conducted using the procedures and guidance endorsed by Regulatory Guide (RG) 1.160, Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Revision 4 (ML18220B281). RG 1.160 endorses the guidance in Section 11 of Nuclear Energy Institute, Nuclear Management Resources Council (NUMARC), NUMARC 93-01, Revision 4F, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, (ML18120A069).

The NRC staff reviewed the TSTF601 information about entering the Applicability of an LCO with the LCO not met. The NRC staff notes that the STS allow continued operation with the shield building inoperable in Mode 1 for the duration of the Completion Time. Since this is allowable, and since in general the risk impact in that particular Mode bounds the risk of transitioning into and through the applicable Modes in the Applicability of the LCO, the use of the LCO 3.0.4.b allowance would be acceptable, as long as the risk is assessed and managed as stated above.

Therefore, based on the above, the NRC staff finds the proposed new Condition A is acceptable in part because of the requirement to perform a risk assessment (LCO 3.0.4.b) to determine the acceptability of entering the Applicability for the shield building LCO with the LCO not met.

3.1.5 Summary Based on the information provided in TSTF601 and the NRC staff evaluation above (SE Sections 3.1.1 through 3.1.4), the NRC staff concludes that the proposed Action (new Condition A) is reasonable given the differences in the plant conditions under the proposed Action when compared to full power. In particular, when operating at full power the current STS Bases describes 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as a reasonable Completion Time in the event shield building operability is not maintained considering the limited leakage design of containment and the low probability of a design-basis accident occurring during this time period. NRC staff considers 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> as a reasonable Completion Time for the following reasons: (1) the limited leakage design of containment, (2) the low probability of a design-basis accident occurring during this same time period, and (3) the lower energy conditions in the reactor resulting in a significant reduction in radionuclide transport coupled with the reduction in radionuclide inventory available for release in Modes 4 and 3 following a refueling and before criticality.

The regulation at 10 CFR 50.36(c)(2) states, in part, that when an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TS until the condition can be met. The proposed change provides a remedial action to be taken when the LCO is not met, and, therefore, the NRC staff finds that 10 CFR 50.36(c)(2) will continue to be met.

3.2 Evaluation of Changes to Existing Actions The proposed change renamed the existing Condition A, which states, Shield Building inoperable, to Condition B and modified it to state, Shield building inoperable for reasons other than Condition A. The renamed Condition B Required Action and Completion Time are unaffected. Existing Action B is renamed Action C but otherwise unchanged.

The NRC staff finds these changes acceptable because they are conforming changes resulting from the new Condition A and do not alter the way the TS are implemented when the shield building is inoperable for reasons other than Condition A.

3.3 STS Change Consistency The NRC staff reviewed the proposed STS changes for technical clarity and consistency with the existing requirements for customary terminology and formatting. The wording of the new Action for the shield building is modeled on an existing allowance (Action A) in the NUREG-1431 and NUREG1432 specifications for the auxiliary feedwater (AFW) system, which permits a longer Completion Time for an inoperable turbine driven AFW pump in Mode 3 following refueling if Mode 2 has not been entered. The NRC staff concludes that the TSTF601 proposed change to the shield building STS is consistent with existing STS requirements for customary terminology and formatting in accordance with SRP Chapter 16.0 and is, therefore, acceptable.

3.4 Consideration of Changes to the STS Bases The NRC staff reviewed the TSTF601, Revision 1, proposed changes to the STS Bases for NUREG-1431 and NUREG1432. As discussed in Section 2.0 of this SE, the Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors describes the scope and purpose of the STS Bases. It does so by listing five questions the STS Bases must address. While the STS Bases must address these questions, not every question will be relevant to every change to the STS Bases. The second, fourth, and fifth questions are not relevant to this evaluation because the STS changes proposed in TSTF601, Revision 1, as evaluated above, do not affect the LCO or its Applicability Bases, safety limits, or surveillance requirements. In addition, because the proposed change only affects Actions, the Policy Statement criterion that applies to the LCO is not affected and the first question is not relevant to this evaluation. The proposed STS Bases support new and revised action statements; therefore, only the third question is relevant to the changes.

Traveler TSTF601 Bases, in part, describe that new Action A allows the shield building to be inoperable for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> instead of the 24-hour Completion Time in Condition B. In addition, the Bases clarified that the 72-hour Completion Time was reasonable in MODE 3 or 4 immediately following a refueling when the reactor has not been critical because of the reduced decay heat generation and radionuclide inventory, the limited leakage design of the containment, the Operable TS systems in the applicable Modes, and the low probability of an event requiring the shield building.

The NRC staff finds the proposed STS Bases changes to be acceptable because the Bases for the new and revised action statements adequately address Question 3 of the Final Policy Statement on TS, ensuring that 10 CFR 50.36 will continue to be met.

3.5 Reportability Traveler TSTF601 Section 3.6, Reportability, referenced information described within NUREG1022, Event Report Guidelines, 10 CFR 50.72 and 50.73. The NRC staff considers this information to be outside the scope of the TSTF601 review. As such, the NRC staff did not review or evaluate the information regarding event report guidelines provided in TSTF601 Section 3.6.

4.0 CONCLUSION

The NRC staff concludes that there is reasonable assurance that plants adopting TSTF-601 will continue to ensure that when the shield building LCO is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TS until the condition can be met.

Therefore, the NRC staff finds that the proposed changes to the STS are acceptable because they continue to meet the requirements of 10 CFR 50.36(c)(2) and provide protection to the health and safety of the public.

Principal Contributors:

Clint Ashley, NRR/DSS/STSB Steve Smith, NRR/DSS/STS Sean Meighan, NRR/DRA/ARCB Ahsan Salman, NRR Derek Scully, NRR/DSS/SCPB Date: September 8, 2025