ML20141L612

Issuance of Amendment No. 219 to Revise Technical Specifications to Reduce Safety Injection Accumulators Minimum Pressures (Exigent Circumstances)

ML20141L612
Person / Time
Site:	South Texas
Issue date:	05/28/2020
From:	Dennis Galvin Plant Licensing Branch IV
To:	Gerry Powell South Texas
Galvin D
References
EPID L-2020-LLA-0108
Download: ML20141L612 (18)

Text

May 28, 2020 Mr. G. T. Powell President and Chief Executive Officer STP Nuclear Operating Company P.O. Box 289 Wadsworth, TX 77483

SUBJECT:

SOUTH TEXAS PROJECT, UNIT 1 - ISSUANCE OF AMENDMENT NO. 219 TO REVISE TECHNICAL SPECIFICATIONS TO REDUCE SAFETY INJECTION ACCUMULATORS MINIMUM PRESSURES (EXIGENT CIRCUMSTANCES)

(EPID L-2020-LLA-0108)

Dear Mr. Powell:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 219 to Renewed Facility Operating License No. NPF-76 for South Texas Project, Unit 1. The amendment consist of changes to the technical specifications (TSs) in response to your application dated May 13, 2020.

The one-time amendment modifies TS 3/4.5.1, Accumulators, to allow Unit 1 to operate with all three safety injection accumulators at reduced minimum pressure for the remainder of the current Unit 1 operating cycle, Cycle 23.

A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions biweekly Federal Register notice.

Sincerely,

/RA/

Dennis J. Galvin, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-498

Enclosures:

1. Amendment No. 219 to NPF-76

2. Safety Evaluation cc: Listserv

STP NUCLEAR OPERATING COMPANY DOCKET NO. 50-498 SOUTH TEXAS PROJECT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 219 Renewed License No. NPF-76

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by STP Nuclear Operating Company (STPNOC)*, acting on behalf of itself and for NRG South Texas LP, the City Public Service Board of San Antonio (CPS), and the City of Austin, Texas (COA)

(the licensees), dated May 13, 2020, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

• STPNOC is authorized to act for NRG South Texas LP, the City Public Service Board of San Antonio, and the City of Austin, Texas, and has exclusive responsibility and control over the physical construction, operation, and maintenance of the facility.

Enclosure 1

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-76 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 219, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. STPNOC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. The license amendment is effective as of its date of issuance and shall be implemented within 14 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Thomas J. Thomas J. Wengert Date: 2020.05.28 Wengert 11:31:07 -04'00' Jennifer L Dixon-Herrity, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License No. NPF-76 and Technical Specifications Date of Issuance: May 28, 2020

ATTACHMENT TO LICENSE AMENDMENT NO. 219 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-76 SOUTH TEXAS PROJECT, UNIT 1 DOCKET NO. 50-498 Replace the following pages of the Renewed Facility Operating License No. NPF-76, and Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Renewed Facility Operating License No. NPF-76 REMOVE INSERT Technical Specifications REMOVE INSERT 3/4 5-1 3/4 5-1

SOUTH TEXAS RENEWED LICENSE (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 219, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. STPNOC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) Not Used (4) Initial Startup Test Program (Section 14, SER)*

Any changes to the Initial Test Program described in Section 14 of the Final Safety Analysis Report made in accordance with the provisions of 10 CFR 50.59 shall be reported in accordance with 50.59(b) within one month of such change.

(5) Safety Parameter Display System (Section 18, SSER No. 4)*

Before startup after the first refueling outage, HL&P[**] shall perform the necessary activities, provide acceptable responses, and implement all proposed corrective actions related to issues as described in Section 18.2 of SER Supplement 4.

(6) Supplementary Containment Purge Isolation (Section 11.5, SSER No. 4)*

HL&P shall provide, prior to startup from the first refueling outage, control room indication of the normal and supplemental containment purge sample line isolation valve position.

• The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.

• • The original licensee authorized to possess, use and operate the facility was HL&P. Consequently, historical references to certain obligations of HL&P remain in the license conditions.

Renewed License No. NPF-76 Amendment No. 219

3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.5.1 Each Safety Injection System accumulator shall be OPERABLE APPLICABILITY: MODES 1 and 2 MODE 3 with pressurizer pressure > 1000 psig ACTION:

a. With one accumulator inoperable, except as a result of boron concentration outside the required limits, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> restore the inoperable accumulator to OPERABLE status or apply the requirements of the CRMP, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b. With more than one accumulator inoperable, except as a result boron concentration outside the required limits, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore at least two accumulators to OPERABLE status or apply the requirements of the CRMP, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

c. With the boron concentration of one accumulator outside the required limit, within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> restore the boron concentration to within the required limits or apply the requirements of the CRMP, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

d. With the boron concentrations of more than one accumulator outside the required limit, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore the boron concentration of at least two accumulators to within the required limits or apply the requirements of the CRMP, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.5.1.1 Each accumulator shall be demonstrated OPERABLE:

a. At a frequency in accordance with the Surveillance Frequency Control Program by:

1) Verifying the contained borated water volume is 8800 gallons and 9100 gallons and nitrogen cover-pressure is 590 psig1 and 670 psig, and

2) Verifying that each accumulator isolation valve is open.

b. At a frequency in accordance with the Surveillance Frequency Control Program and within 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />s* after each solution volume increase of greater than or equal to 1% of tank volume that is not the result of addition from the RWST by verifying the boron concentration of the accumulator solution is 2700 ppm and 3000 ppm, and

c. At a frequency in accordance with the Surveillance Frequency Control Program when the RCS pressure is above 1000 psig by verifying that power to the isolation valve operator is removed.

• The 6 hr. SR is only required to be performed for affected accumulators.

1 For Unit 1 only, the nitrogen cover-pressure is verified to be 500 psig for the remainder of Unit 1 Cycle 23.

SOUTH TEXAS - UNITS 1 & 2 3/4 5-1 Unit 1 Amendment No. 51,54,59,135 179 188 219 Unit 2 Amendment No. 40.43.47,124 166 175

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 219 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-76 STP NUCLEAR OPERATING COMPANY, ET AL.

SOUTH TEXAS PROJECT, UNIT 1 DOCKET NO. 50-498

1.0 INTRODUCTION

By letter dated May 13, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20134K758), STP Nuclear Operating Company (STPNOC, the licensee) submitted a license amendment request (LAR) to the U.S. Nuclear Regulatory Commission (NRC, the Commission) for South Texas Project (STP), Unit 1.

The proposed one-time amendment would modify STP Technical Specification (TS) 3/4.5.1, Accumulators, to allow Unit 1 to operate with all three safety injection (SI) accumulators at reduced minimum pressure for the remainder of the current Unit 1 operating cycle (Cycle 23).

Specifically, a footnote would be added to TS Surveillance Requirement (SR) 4.5.1.1.a.1) to reduce the lower pressure limit for the accumulator nitrogen cover-pressure from 590 pounds per square inch gauge (psig) to 500 psig during Unit 1 Cycle 23.

Pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 50.91(a)(6), the licensee requested that the proposed amendment be issued under exigent circumstances.

A detailed discussion of the exigent circumstances is contained in Section 4.0 of this safety evaluation (SE).

2.0 REGULATORY EVALUATION

2.1 System Description Accumulators are pressure vessels filled with borated water and pressurized with nitrogen gas that function to supply water to the reactor vessel during a loss of coolant accident (LOCA).

They are passive safety system components that activate when the reactor vessel pressure drops below the accumulator pressure, piping water into the reactor coolant system (RCS) cold leg through an accumulator line. STP Unit 1 has three emergency core cooling system SI accumulators. Each accumulator is isolated from the RCS by two check valves in series.

TS 3.5.1 states that each SI system accumulator is required to be operable in Operational Enclosure 2

Modes 1, 2, and in Mode 3 with pressurizer pressure greater than 1000 psig. The safety function of the accumulators is described in Section 6.3.2, System Design, of the STP Updated Final Safety Analysis Report (UFSAR).

2.2 Description of Proposed TS Change SR 4.5.1.1.a.1) requires demonstrating accumulator operability by verifying the contained borated water volume is greater than or equal to () 8800 gallons and less than or equal to

() 9100 gallons and nitrogen cover-pressure is 590 psig and 670 psig.

The licensee proposed adding a footnote to SR 4.5.1.1.a.1) to reduce the minimum allowed nitrogen cover-pressure for the Unit 1 accumulators to 500 psig for the remainder of Unit 1 Cycle 23. The footnote would state:

1 For Unit 1 only, the nitrogen cover-pressure is verified to be 500 psig for the remainder of Unit 1 Cycle 23.

2.3 Regulatory Requirements The regulation in 10 CFR 50.36(b) states, in part, that Each license authorizing operation of a production or 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 § 50.34 [(describing the technical information to be included in applications for an operating license)].

Pursuant to 10 CFR 50.36(c), TSs are required to include items in the following five specific categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) SRs; (4) design features; and (5) administrative controls. The Commission may include such additional TSs as the Commission finds appropriate.

Paragraph 50.36(c)(3) of 10 CFR establishes the requirements for SRs and states:

Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

Section 50.46 of 10 CFR, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, mandates acceptable criteria for emergency core cooling systems for light-water nuclear power reactors.

The following general design criteria (GDC) in 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, Appendix A, General Design Criteria for Nuclear Power Plants, are applicable to this review.

GDC 4, Environmental and dynamic effects design bases, which requires structures, systems, and components important to safety be protected against dynamic effects associated with flow instabilities and loads such as those resulting from water hammer.

GDC 27, Combined reactivity control system capability, which requires the reactivity control systems to be designed to have a combined capability, in conjunction with poison addition by the emergency core cooling system, of reliably controlling reactivity changes under postulated accident conditions.

GDC 35, Emergency core cooling, which requires a system to provide abundant emergency core cooling to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented, and (2) clad metal-water reaction is limited to negligible amounts.

GDC-50, Containment design basis, insofar as it requires that the containment and its associated heat removal systems be designed so that the containment structure can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated temperature and pressure conditions resulting from any LOCA.

3.0 TECHNICAL EVALUATION

The regulatory framework the NRC staff used to determine the acceptability of the proposed changes consist of the requirements listed in Section 2.3 of this SE. The NRC staff reviewed the proposed TS changes to determine whether they meet the standards in 10 CFR 50.36 and whether the requirements of 10 CFR 50.46 and the applicable GDC will continue to be met.

The licensee provided evaluations and analyses to justify the proposed changes to SR 4.5.1.1.a.1) in Sections 3 and 4 of the LAR.

3.1 Background

During startup from the most recent Unit 1 refueling outage (1RE22), the licensee identified leakage through check valve RH-0032A. This is the low head safety injection (LHSI) Train A to Loop 1A cold leg check valve. The leakage rate was identified as 6 gallons per hour, or 144 gallons per day. Excessive leakage through check valve RH-0032A transports accumulator water to the residual heat removal (RHR) system through the Train A RHR discharge header.

The excess accumulator leakage through check valve RH-0032A will lead to the RHR and LHSI systems exceeding the relief valve pressure setpoints. To remedy this problem, the licensee must perform manipulations of various safety-related components to vent the RHR system and to refill SI Accumulator 1A. Currently, the licensee has isolated SI Accumulator 1A and has entered TS 3.5.1 Action a for which the risk-informed completion time ends May 29, 2020.

3.2 Proposed TS Change The licensee requests a one-time TS change, which would allow the accumulator nitrogen cover-pressure to be 500 psig for the remainder of Unit 1 Operating Cycle 23. Currently, TS 3.5.1 SR 4.5.1.1.a.1) requires the nitrogen cover-pressure to be 590 psig and 670 psig.

The TS will be modified by a footnote as described in Section 2.2 of this SE.

3.3 Evaluation of Changes to Accident Analyses Large Break Loss of Coolant Accident The limiting value of interest for a large-break LOCA (LBLOCA) is the peak clad temperature (PCT). The licensee evaluated the effect of the decrease in accumulator pressure to 500 psig.

As defined in 10 CFR 50.46, the acceptance criteria limiting value for PCT is 2200 degrees Fahrenheit (°F). The current analysis of record (AOR) for STP is 2123 °F for Unit 1 and the PCT with the proposed reduced accumulator pressure for Unit 1 is 2126 °F. The increase in PCT can be attributed to a delay in accumulator injection that accompanies the lowered accumulator pressure. The licensee evaluated the accumulator pressure change by using the Westinghouse LBLOCA evaluation model with BASH (BASH-EM), approved for use in 1981.

This is the current approved analysis methodology for STP, which the licensee uses for its AOR.

The licensees method to evaluate the effects of fuel pellet thermal conductivity degradation on PCT is the LOCBART Transient Extension Method (LTEM). The licensee complied with all NRC conditions and limitations found in the NRC SEs for these methodologies. The licensee determined the PCT change by evaluating only the limiting case for PCT. The NRC staff has determined that only the limiting PCT LBLOCA case needs to be evaluated because a change in accumulator pressure does not impact the overall thermal-hydraulic response during a LBLOCA. The change of 3 °F resulting from the accumulators reduced nitrogen cover pressure is minor and continues to allow for adequate protection, as there continues to be sufficient margin to the 10 CFR 50.46 limit. The NRC staff finds the results of this evaluation, using the previously-approved methodology, to be acceptable.

Small Break Loss of Coolant Accident The STP limiting small break LOCA (SBLOCA) is the 2.0-inch break. This accident does not rely on accumulator injection for mitigation. There is no accumulator injection because the RCS pressure remains above the accumulator injection pressure throughout the entirety of the accident scenario. The licensee evaluated additional break sizes to determine what size break would require accumulator injection for mitigation. The licensee determined that any break less than 4.0 inches would not require accumulator injection. Since the STP limiting break for SBLOCA is 2.0 inches and does not require accumulator injection for mitigation, the AOR remains valid and re-evaluation with accumulator cover gas pressure at 500 psig is not required.

The NRC staff finds this acceptable.

Containment Analysis The function of the containment is to limit offsite radiation dose levels. Offsite radiation dose levels are impacted by containment leakage. To protect against containment leakage, a containment analysis is performed to the peak pressure, temperature and temperature profile following design-basis accidents (DBAs). The limiting DBA for containment pressure is the LBLOCA and the limiting DBA for containment temperature and temperature profile, used for safety-grade equipment qualification, is the main steam line break (MSLB).

The containment mass and energy release analysis was performed for the LBLOCA with an accumulator cover-gas pressure of 500 psig. This analysis resulted in an increase of peak containment pressure by 0.2 pounds per square inch and delayed the accumulator injection by 2 seconds. The current AOR for STP Unit 1 is 40.1 psig. Peak containment pressure can also affect the containment integrated leak rate test (ILRT), discussed in Section 6.2.6.1, Containment Integrated Leakage Rate Test (Type A), of the STP UFSAR. The maximum

containment pressure for ILRT is specified in TS Bases 3/4.6.2.1 and is 41.2 psig. Since the containment design pressure is 56.5 psig, and the containment (ILRT) pressure is 41.2 psig, the increase of peak containment pressure to 40.3 psig is minor, below AOR limits, and continues to meet the requirements set forth in GDC 50. The NRC staff finds this change to be acceptable.

During the MSLB, the RCS pressure remains high throughout the transient. The requested change in accumulator pressure does not impact the MSLB analysis results because the accumulators do not inject during an MSLB. Therefore, the containment peak temperature is unchanged. The NRC staff finds this acceptable.

Electrical components of safety-related equipment are qualified for their normal operational environment and the worst-case DBA environment (see UFSAR Section 6.2.1.1.3.8, Equipment Qualification). The worst-case DBA equipment is qualified for the LOCA and MSLB. The design criteria for equipment qualification that the licensee uses for analysis are 45.0 psig and 330 °F. As the calculated values resulting from the proposed accumulator pressure change are less than the design criteria values, the current STP analysis for equipment qualification remains valid. The NRC staff finds this acceptable.

The containment subcompartment integrity analysis must be evaluated for any impact as a result of lowering the accumulator pressure. The analysis assumes a double-ended rupture of the RHR piping between the hot leg piping and the first isolation valve in the 12-inch nominal diameter section of the RHR piping. The NRC staff has previously approved the licensees use of a leak-before-break methodology, which excludes 8-inch, 10-inch, and 12-inch accumulator lines. This methodology includes the portion of the RHR lines up to the first isolation valve closest to the RCS. Since the accumulator lines are excluded, a change in accumulator cover-gas pressure does not impact the containment subcompartment analysis. The NRC staff finds this acceptable.

Based on the evaluations addressed above, the NRC staff finds that STP Unit 1 continues to meet its containment design basis and therefore meets the requirements set forth in GDC-50.

LOCA Dose Analysis A change to the LBLOCA AOR from the accumulator pressure change can impact the dose analysis. The dose is impacted if there is an increase in radio-isotope inventory release from the fuel, or if there is a change in mass and energy releases as a result of an increase in containment pressure and peak sump water temperature. The licensee evaluated the impact on dose due to these changes. The licensee found no change in radioisotope inventory release due to the increase in the LBLOCA PCT because the result remains below the 10 CFR 50.46 acceptance criteria. The containment pressure change also does not impact the dose analysis because the peak pressure remains below acceptance limits. Finally, the licensee calculated a peak sump water temperature increase in the LBLOCA analysis. The peak sump water increases from 274.1 °F to 274.3 °F due to the change in accumulator pressure. The peak sump water temperature can impact the sump pH analysis for the elemental iodine decontamination factor used in the LBLOCA dose analysis. However, the licensee used a peak sump water temperature of 275 °F in its AOR. The LBLOCA dose analysis is not impacted because the AOR is bounding. The NRC staff finds this acceptable.

Post-LOCA Long Term Cooling The post-LOCA long-term cooling analysis does not use accumulator cover-gas pressure as an input. The post-LOCA long-term cooling analysis also includes post-LOCA decay heat removal, subcriticality checks, and boric acid precipitation control calculations to determine the time hot leg recirculation should be initiated. The AOR is not impacted by a change in accumulator cover-gas pressure and therefore, remains valid. The NRC staff find this acceptable.

Other Transients The NRC staff confirmed that additional STP UFSAR Chapter 15 transients are not mitigated by accumulator injection for Unit 1. Therefore, a change in accumulator pressure to 500 psig does not impact the AOR. The NRC staff finds this acceptable.

3.4 Gas Accumulation Due to check valve RH-0032A leakage, the LHSI/RHR header can potentially accumulate nitrogen gas coming from the accumulator. The licensee used its Gas Accumulation Management Plan to establish a monitoring plan and document the boundary check valve. The Gas Accumulation Management Plan establishes monitoring frequency and acceptance criteria for gas accumulation in piping configurations.

The licensee evaluated three likely scenarios for gas accumulation. The first scenario is gas coming out of solution in the LHSI/RHR discharge header during LOCA depressurization prior to LHSI pump start. The second scenario is gas out of solution upstream of check valve SI-0030A, the new pressure boundary. The third scenario is gas out of solution when the RHR system is depressurized for situations where the RHR system is used for normal plant cooldown, post-accident cooldown, and RHR inservice surveillance testing.

In scenario one for a LBLOCA, the concern is that a water hammer could develop because gas voids will coalesce into larger voids, and the gas volume will be homogenous or stratified with the liquid phase prior to pump start. There is no downstream pressure boundary in this scenario because the injection check valves will be open as a result of the LBLOCA. Therefore, a water hammer will not occur when the LHSI pump starts because the pressure wave will be dissipated in the RCS.

During an SBLOCA in scenario one, the LHSI pump will already be in operation and the LHSI/RHR header will remain filled with water until RCS pressure drops to the injection point.

Gas voids are expected to be small and will not impact injection time.

In scenario two, which involves a void upstream of check valve SI-0030A, a void compression can occur if the RCS pressure remains high, but it is unlikely that the void would travel to a lower elevation pump suction. This scenario would not impact RCS injection.

Finally, in scenario three, there are two potential situations. First, if the gas void is large enough, it can cause inertial deceleration of an accelerating water column. The licensee has determined that this case will result in an oscillating pressure response, but no propagating water hammer event. The second situation involves a gas volume small enough that the accelerating water column impacts the fluid on the other side of the void. While the licensee has determined that this situation could lead to propagating water hammer pressures, those pressures are not large enough to impact operation of the RHR system negatively.

The NRC staff has determined that the licensee has adequately evaluated the potential impact of gas accumulation and will maintain compliance with the requirements set forth in GDC 4.

3.5 Evaluation of System Configuration With leakage through check valve RH-0032A, the accumulator will be at the same pressure as the LHSI/RHR header and will be, in effect, one system. In this configuration, the LHSI/RHR header will be pressurized to 500 psig. This pressure is acceptable in the header because pressure relief in the header occurs at 600 psig, as shown in the schematic included in the LAR submittal. Additionally, the water in the LHSI/RHR header is maintained at the same boron concentration as the accumulators because the header is supplied by the reactor water storage tank (RWST), which contains borated water. Section 6.3.2.2, Equipment and Component Descriptions, of the STP UFSAR states, in part, that accumulator water level can be adjusted by pumping borated water in from the RWST. Input parameters described in Section 6.5, Fission Product Removal and Control Systems, of the STP UFSAR show the boron concentration in the RWST to be equivalent to the accumulator boron concentration. The NRC staff has determined that the requirements set forth in GDC 27 will continue to be met for this configuration.

3.6 Technical Conclusions The NRC staff determined that STP TS SR 4.5.1.1.a.1), as amended by the footnote specific to Unit 1, will continue to provide an acceptable way to meet the requirements of 10 CFR 50.36(c)(3), 10 CFR 50.46, and GDC 4, 27, 35, and 50, because the revised SR will continue to provide assurance that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

4.0 EXIGENT CIRCUMSTANCES

4.1 Background

The NRCs regulations contain provisions for the issuance of amendments when the usual 30-day prior public comment period cannot be met. These provisions are applicable under exigent circumstances. Consistent with the requirements in 10 CFR 50.91(a)(6), exigent circumstances exist when: (1) a licensee and the NRC must act quickly; (2) time does not permit the NRC to publish a Federal Register notice allowing 30 days for prior public comment; and (3) the NRC determines that the amendment involves no significant hazards consideration.

As discussed in the licensees application dated May 13, 2020, the licensee requested that the proposed amendment be processed by the NRC on an exigent basis.

Under the provisions in 10 CFR 50.91(a)(6), the NRC notifies the public in one of two ways when exigent circumstances exist: (1) by issuing a Federal Register notice providing an opportunity for hearing and allowing at least 2 weeks from the date of the notice for prior public comments; or (2) by using local media to provide reasonable notice to the public in the area surrounding the licensees facility. In this case, the NRC used local media and published a public notice in The Bay City Tribune located in Bay City, Texas (https://baycitytribune.com/),

a newspaper local to the licensees facility, on May 20, 2020.

4.2 The Licensees Basis for Exigency The licensee provided the following information to explain the exigency of the proposed amendment. The licensee stated that the emergent condition necessitating this request for exigency was caused by increased valve leakage during startup from the Unit 1 refueling outage. The licensee identified a significant increase in leakage through check valve RH-0032A, the LHSI Train A to Loop 1A cold leg check valve, during startup from Unit 1 refueling outage 1RE22. The licensee identified that SI Accumulator 1A was losing inventory at approximately 6 gallons per hour (144 gallons per day) through check valve RH-0032A, which was transporting accumulator water to the RHR system through the Train A RHR discharge header and causing the header to pressurize. The licensee stated that it cannot maintain the RHR and LHSI systems below the relief valve setpoints without performing excessive manipulations of safety-related components required to vent the RHR system and refill SI Accumulator 1A. These actions could lead to equipment failure and an unplanned shutdown.

Alternatively, a reactor shutdown would be required to make repairs.

After considering several options to address the condition, the licensee determined that the safest solution was to isolate Unit 1 SI Accumulator 1A and then request approval of a TS change to reduce the minimum allowed cover-gas pressure for the accumulators prior to restoration. Per the requirements of STP TS 6.8.3.k, Configuration Risk Management Program (CRMP), Unit 1 can only remain in this configuration for 30 days, which is the backstop risk-informed completion time allowed by the TS. Because STP Unit 1 will reach the backstop risk-informed completion time on May 29, 2020, the licensee determined that the need for this LAR is exigent and does not allow for the standard public comment period. Based on this information, the NRC staff finds that exigent circumstances exist in that the licensee and the NRC must act quickly and that time does not permit the NRC staff to publish a Federal Register notice allowing 30 days for prior public comment.

4.3 NRC Staff Conclusion

Based on the above circumstances, the NRC staff finds that the licensee made a timely application for the proposed amendment following its identification of the issue. In addition, the NRC staff finds that timely action is required to prevent a reactor shutdown or excessive manipulations of safety-related components. Based on these findings, and the determination that the amendment involves no significant hazards consideration as discussed in Section 6.0 below, the NRC staff has determined that a valid need exists for issuance of the license amendment using the exigent provisions of 10 CFR 50.91(a)(6).

5.0 PUBLIC COMMENTS Under the provisions in 10 CFR 50.91(a)(6), the NRC used local media and published a public notice in the The Bay City Tribune located in Bay City, Texas (https://baycitytribune.com/), a newspaper local to the licensees facility, on May 20, 2020. The notice included the NRC staffs proposed no significant hazards consideration determination. The notice also provided an opportunity for public comment until May 27, 2020, regarding the NRC staffs proposed no significant hazards consideration determination. No public comments were received.

6.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION

The NRCs regulation in 10 CFR 50.92(c) states that the NRC may make a final determination, under the procedures in 10 CFR 50.91, that a license amendment involves no significant

hazards consideration if operation of the facility, in accordance with the amendment, would not:

(1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

As required by 10 CFR 50.91(a), in its application dated May 13, 2020, the licensee provided its analysis of the issue of no significant hazards consideration, which is presented below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change reducing the lower pressure limit for the accumulators does not affect the design basis of the plant - this change will have no significant impact on the capability of the accumulators to perform their design basis function. The overall protection system performance will remain within the bounds of the accident analyses documented in Chapter 15 of the STP Updated Final Safety Analysis Report (UFSAR). The change does not affect other structures, systems and components of the plant or change plant operations, design functions or analysis that verifies design functions. The accumulators are not a precursor to any accident previously evaluated. The change results in an insignificant increase in the consequences of the safety analyses and STP Unit 1 will remain within licensing basis limits. The accumulators will continue to function in a manner consistent with safety analyses assumptions and the plant design basis. There will be no degradation in the performance of, or an increase in the number of challenges to, equipment assumed to function during any postulated accident.

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

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change does not involve a physical alteration to the plant or a change in the methods governing normal plant operation - no new or different type of equipment will be installed. The proposed change does not adversely affect the design function or operation of any structures, systems, or components important to safety. No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures are introduced because of the proposed change. Safety systems continue to function in the same manner and there is no reliance on additional systems or procedures. The accumulators are passive components that are not accident initiators. Lowering the accumulator pressure would not create a new or different kind of accident. The malfunction of safety-related equipment, assumed to be operable in the accident analyses,

would not result from the proposed change. No new failure mode has been created and no new equipment performance burdens are imposed.

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

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change affects the margins related to Peak Clad Temperature, Containment Peak Pressure, and Containment Equipment Qualification Peak Pressure. The analysis of these parameters determined that the margins to the design basis and safety limits were slightly reduced, however the design bases and safety limits were not impacted. The performance of the accumulators remains within design basis limits and the accident analyses safety limits.

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

The NRC staff reviewed the licensees no significant hazards consideration analysis. Based on this review and on the NRC staffs evaluation of the underlying LAR as discussed above, the NRC staff concludes that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff makes a final determination that no significant hazards consideration is involved for the proposed amendment and that the amendment should be issued as allowed by the criteria contained in 10 CFR 50.91.

7.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Texas State official was notified of the proposed issuance of the amendment on May 20, 2020. The State official had no comments.

8.0 ENVIRONMENTAL CONSIDERATION

The amendment changes requirements with respect to installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and changes SRs. 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 made a final determination that no significant hazards consideration is involved for the proposed amendment as discussed above in Section 6.0 of this SE. 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.

9.0 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) there is reasonable assurance that such activities will be conducted in compliance with the Commissions 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.

Principal Contributors: D. Woodyatt M. Hamm Date: May 28, 2020

ML20141L612 *via email OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA* NRR/DSS/SNSB/BC* NRR/DSS/STSB/BC*

NAME DGalvin PBlechman SKrepel VCusumano DATE 5/21/20 5/21/20 5/20/20 5/20/20 OFFICE OGC* NLO NRR/DORL/LPL4/BC* NRR/DORL/LPL4/PM*

JDixon-Herrity (TWengert NAME STurk for) DGalvin DATE 5/21/20 5/28/20 5/28/20