Text

Issuance of Amendment Nos. 227 and 212 to Authorize the Revision of the Alternative Source Term Dose Calculation
	ML24022A225
Person / Time
Site:	South Texas
Issue date:	02/20/2024
From:	Dennis Galvin; Plant Licensing Branch IV
To:	Gerry Powell; South Texas
	Galvin D
References
	EPID L-2023-LLA-0047
	Download: ML24022A225 (23)
	v • d • e

February 20, 2024 Mr. G. T. Powell President and Chief Executive Officer STP Nuclear Operating Company P.O. Box 289 Wadsworth, TX 77483

SUBJECT:

SOUTH TEXAS PROJECT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 227 AND 212 TO AUTHORIZE THE REVISION OF THE ALTERNATIVE SOURCE TERM DOSE CALCULATION (EPID L-2023-LLA-0047)

Dear Mr. Powell:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 227 to Renewed Facility Operating License No. NPF-76 and Amendment No. 212 to Renewed Facility Operating License No. NPF-80 for the South Texas Project, Units 1 and 2, respectively. The amendments authorize revision to the Updated Final Safety Analysis Report (UFSAR) in response to your application dated March 30, 2023, as supplemented by letters dated October 19, 2023, and February 8, 2024.

The amendments authorize revision to the UFSAR to revise the alternative source term dose calculation for the main steam line break and the locked rotor accident. The reanalysis uses the asymmetric natural circulation cooldown thermohydraulic analyses, various radiation transport assumptions, and the current licensing basis source term and meteorological data to evaluate the dose effects of an extended cooldown on the existing accident analyses.

A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions monthly 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 Nos. 50-498 and 50-499

Enclosures:

1. Amendment No. 227 to NPF-76

2. Amendment No. 212 to NPF-80

3. 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. 227 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 Constellation South Texas, LLC, the City Public Service Board of San Antonio (CPS), and the City of Austin, Texas (COA) (the licensees), dated March 30, 2023, as supplemented by letters dated October 19, 2023, and February 8, 2024, 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 Constellation South Texas, LLC, 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.

2.

Accordingly, by Amendment No. 227, the license is amended to authorize revision to the Updated Final Safety Analysis Report (UFSAR), as set forth in the application dated March 30, 2023, as supplemented by letters dated October 19, 2023, and February 8, 2024. The licensee shall update the UFSAR to incorporate the changes as described in the licensees application dated March 30, 2023, as supplemented by letters dated October 19, 2023, and February 8, 2024, and the NRC staffs safety evaluation enclosed with this amendment and shall submit the revised description authorized by this amendment with the next update of the UFSAR.

3.

This license amendment is effective as of its date of issuance and shall be implemented by October 1, 2024. The UFSAR changes shall be implemented in the next periodic update to the UFSAR in accordance with 10 CFR 50.71(e).

FOR THE NUCLEAR REGULATORY COMMISSION Jennivine K. Rankin, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: February 20, 2024 Jennivine K.

Rankin Digitally signed by Jennivine K. Rankin Date: 2024.02.20 11:38:18 -05'00'

STP NUCLEAR OPERATING COMPANY DOCKET NO. 50-499 SOUTH TEXAS PROJECT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 212 Renewed License No. NPF-80

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 Constellation South Texas, LLC, the City Public Service Board of San Antonio (CPS), and the City of Austin, Texas (COA) (the licensees), dated March 30, 2023, as supplemented by letters dated October 19, 2023, and February 8, 2024, 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 Constellation South Texas, LLC, 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.

2.

Accordingly, by Amendment No. 212, the license is amended to authorize revision to the Updated Final Safety Analysis Report (UFSAR), as set forth in the application dated March 30, 2023, as supplemented by letters dated October 19, 2023, and February 8, 2024. The licensee shall update the UFSAR to incorporate the changes as described in the licensees application dated March 30, 2023, as supplemented by letters dated October 19, 2023, and February 8, 2024, and the NRC staffs safety evaluation enclosed with this amendment and shall submit the revised description authorized by this amendment with the next update of the UFSAR.

3.

This license amendment is effective as of its date of issuance and shall be implemented by October 1, 2024. The UFSAR changes shall be implemented in the next periodic update to the UFSAR in accordance with 10 CFR 50.71(e).

FOR THE NUCLEAR REGULATORY COMMISSION Jennivine K. Rankin, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: February 20, 2024 Jennivine K.

Rankin Digitally signed by Jennivine K. Rankin Date: 2024.02.20 11:38:44 -05'00'

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 227 AND 212 TO RENEWED FACILITY OPERATING LICENSE NOS. NPF-76 AND NPF-80 STP NUCLEAR OPERATING COMPANY, ET AL.

SOUTH TEXAS PROJECT, UNITS 1 AND 2 DOCKET NOS. 50-498 AND 50-499

1.0 INTRODUCTION

By letter dated March 30, 2023 (Reference 1), as supplemented by letters dated October 19, 2023 (Reference 2), and February 8, 2024 (Reference 3), the 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, Units 1 and 2 (STP).

The proposed amendments would request approval to revise the alternative source term (AST) dose calculation for the main steam line break (MSLB) and the locked rotor accident (LRA). The reanalysis uses the asymmetric natural circulation cooldown (ANCC) thermohydraulic analyses, various radiation transport assumptions, and the current licensing basis (CLB) source term and meteorological data to evaluate the dose effects of an extended cooldown on the existing accident analyses. The LAR does not propose any changes to the technical specifications.

The supplemental letters dated October 19, 2023, and February 8, 2024, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register (FR) on June 13, 2023 (88 FR 38552).

2.0 REGULATORY EVALUATION

2.1 Background

The NRC previously approved the application of AST for STP for design-basis accidents (DBAs)

(Reference 4). The licensee recognized that steam generator (SG) loop flow stagnation would occur during an event with ANCC with higher cooldown rates if one or more SGs were not available. To avoid reactor coolant system (RCS) loop flow stagnation, the cooldown rate should be reduced in an ANCC situation. This results in an extended cooldown timeline. The licensee determined that the RCS cooldown rate in these conditions should be limited to a rate

dependent on the temperature differential present in the active loops to prevent SG loop stagnation.

In 2008, the licensee applied the AST analysis methodology by assuming an 8-hour cooldown timeline for many non-loss-of-coolant accident (non-LOCA) DBAs. This was done after the licensee identified the impact of ANCC on the MSLB and LRA events following the reports and operating experiences in similar plants, which reported that industry evaluation of plant cooldown accident analyses involving an inoperable cooling loop found that the cooldown duration can be much longer than assumed.

The licensee acknowledged that the impact of a reduced cooldown rate on a post-accident cooldown timeline for an MSLB identified in Callaway Plant, Unit 1, Licensee Event Report 2018-002-00 (Reference 5) was applicable to STPs MSLB dose analysis. Moreover, the ANCC issue impacts the LRA due to the limiting single failure (loss of engineered safety feature (ESF)

Actuation Signal A). Both MSLB and LRA events are affected by the assumptions applied in this analysis since it requires cooling down an intact RCS without all four SGs during a coincident loss of offsite power (LOOP). Increasing the extended cooldown timeline as a result of an ANCC situation would cause increased dose consequences that exceed more than a minimal increase in the consequences of an accident previously evaluated in the Updated Final Safety Analysis Report (UFSAR) under Title 10 of the Code of Federal Regulations (10 CFR)

Section 50.59(c)(2)(iii).

2.2 Description of Proposed Changes The licensee proposes to revise the radiological dose consequence analyses pursuant to 10 CFR 50.59. The proposed reanalysis of the current design basis AST analysis for MSLB and LRA events is based on the ANCC thermal-hydraulic analysis with higher cooldown rates, conservative radiation transport assumptions, and the CLB source term and meteorological data to evaluate the dose effects of the extended cooldown on the existing accident analyses. The licensee indicated that the analysis was performed according to Regulatory Guide (RG) 1.183 Revision 0, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors (Reference 6). The discussion of RG 1.183 in this safety evaluation (SE) refers to Revision 0 unless otherwise noted.

2.3 Applicable Regulatory Requirements and Guidance The NRC staff evaluated the impact of the proposed changes as they relate to regulations and regulatory guidance.

2.3.1 Regulatory Requirements The NRC staff's evaluation is based upon the following regulations.

The regulation in 10 CFR 50.67, Accident source term, states, in part, in paragraph 50.67(b)(2) that the applicants analysis must demonstrate with reasonable assurance that:

(i)

An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess

of 0.25 Sv [sievert] (25 rem [roentgen equivalent man]) total effective dose equivalent (TEDE).

(ii)

An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).

(iii)

Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident.

Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50, General Design Criterion 10 (GDC 10), Reactor Design, requires in part that the RCS be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

Appendix A to 10 CFR Part 50, GDC 19, Control room, states, in part, that, A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents.

Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident. Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.

Additionally, GDC 19, states that, holders of operating licenses using an alternative source term under

§ 50.67, shall meet the requirements of this criterion, except that with regard to control room access and occupancy, adequate radiation protection shall be provided to ensure that radiation exposures shall not exceed 0.05 Sv (5 rem) total effective dose equivalent (TEDE) as defined in § 50.2 for the duration of the accident.

2.3.2 Regulatory Guidance The NRC staff used the following documents to provide guidance on acceptable approaches to demonstrate that the above regulations are met:

RG 1.183 provides guidance to licensees of operating power reactors on acceptable applications of ASTs; the scope, nature, and documentation of associated analyses and evaluations; consideration of impacts on analyzed risk; acceptable radiological analysis assumptions for use in conjunction with the accepted AST; and content of submittals.

RG 1.183 establishes an acceptable AST and identifies the significant attributes of other ASTs that may be found acceptable by the NRC staff. RG 1.183 also identifies acceptable radiological analysis assumptions for use in conjunction with the accepted AST.

NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition (SRP), Section 15.0.1, Revision 0, Radiological Consequence Analyses Using Alternative Source Terms (Reference 7) provides guidance on reviewing compliance with 10 CFR 50.67, as it relates to the implementation of an AST in current operating nuclear power plants.

3.0 TECHNICAL EVALUATION

3.1 Evaluation of Thermal-Hydraulic Analysis Changes 3.1.1 Description of the Thermal-Hydraulic Analysis To obtain steam release rates during various phases of plant cooldown to the reactor heat removal (RHR) cut-in conditions (i.e. the reduction in the RCS pressure and temperature so that the RHR can provide its heat removal function), the licensee applied the NRC-approved, Westinghouse WCAP-14882-P-A, RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses (Reference 8) methodology based on RETRAN02 software to model the sequence of events after an MSLB or LRA event under ANCC conditions. The steam release rates are then combined with the CLB radiation source term in a SNAP/RADTRAD model described in NUREG/CR-7220, SNAP/RADTRAD 4.0:

Description of Models and Methods (Reference 9) for dose analysis purposes. In a letter dated October 19, 2023, the licensee stated that the proposed RETRAN02 analysis does not replace the CLB thermal-hydraulic analysis for a DBA of an MSLB and LRA. The CLB thermal-hydraulic analysis, described in UFSAR (Reference 10) sections 15.1.5, Spectrum of Steam System Piping Failures Inside and Outside Containment, and 15.3.3, Reactor Coolant Pump Shaft Seizure (Locked Rotor), for an MSLB and LRA, respectively, addresses the core response and departure from nucleate boiling ratio concerns associated with 10 CFR Part 50, Appendix A, GDC 10.

The LAR describes the system design, the operation of the major systems involved in the MSLB and LRA for the thermal-hydraulic analyses, the reasons for the proposed changes, the description of the proposed changes, and detail description of the MSLB and LRA ANCC analyses.

In a letter dated October 19, 2023, the licensee stated, in part, that [t]he assumptions listed in LAR Sections 3.2.1 and 3.3.1 for the RETRAN[02] analysis are intended to get a conservative cooldown timeline and steam release input for the dose analyses. To avoid RCS loop flow

stagnation during an ANCC situation, the licensee determined that the cooldown rates should be limited to 15 degrees Fahrenheit per hour (°F/hr) to 20°F/hr. The RETRAN02 analyses in the LAR for MSLB and LRA under ANCC conditions conservatively assume the plant cools down at the minimum procedurally allowed rate for the entirety of the cooldown process. The licensee stated that in the ANCC situation, the cooldown rates will be controlled by the emergency operating procedure (EOP).

3.1.2 Evaluation of MSLB ANCC Analysis The licensee developed a RETRAN02 model for the MSLB event in the ANCC situation to obtain the extended time to reach RHR cut-in condition and the steam releases. The assumptions used in the MSLB ANCC analysis, the difference between the model and CLB, and the MSLB steam release comparison obtained by the model to the CLB are described in section 3.2.1, MSLB ANCC Cooldown Analysis (Reference 6.1), of the enclosure to the LAR.

Table 1 of this SE shows the comparison between the proposed RETRAN MSLB ANCC analysis results with the current analysis of record. The time units in tables 1 and 2 of this SE are hours (hrs) and the mass units are pound-mass (lbm).

Table 1: MSLB Event Results Comparison with Current Analysis of Record Scenario with MSLB ANCC Current UFSAR Table 15.1-2 RHR cut-in time:

28.0 hrs RHR cut-in time:

8.0 hrs Release from Faulted SG (1 SG):

Release from Faulted SG (1 SG):

0 - 0.55 hours6.365741e-4 days 0.0153 hours 9.093915e-5 weeks 2.09275e-5 months 3.151E+05 lbm 0 - 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 5.990 E+05 lbm Release from Active SGs (2 SGs):

Release from Intact SGs (3 SGs):

0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 8.883E+04 lbm 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 4.520 E+05 lbm 2 - 28 hours3.240741e-4 days 0.00778 hours 4.62963e-5 weeks 1.0654e-5 months 2.578E+06 lbm 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.080 E+06 lbm Release from Inactive SG (1 SG):

0 - 0.02 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.130E+03 lbm 1.29 - 3.49 hours5.671296e-4 days 0.0136 hours 8.101852e-5 weeks 1.86445e-5 months 1.132E+05 lbm 12.54 - 12.57 hours6.597222e-4 days 0.0158 hours 9.424603e-5 weeks 2.16885e-5 months 1.737E+03 lbm 27.44 - 28.00 hours0 days 0 hours 0 weeks 0 months 4.775E+03 lbm Total SG Steam Release:

3.103E+06 lbm Total SG Steam Release:

2.131 E+06 lbm The CLB assumes an 8-hour cooldown timeline to the RHR cut-in condition for MSLB cases, whereas the results of the RETRAN02 analysis determines the RHR cut-in condition to be reached by approximately 28 hours3.240741e-4 days 0.00778 hours 4.62963e-5 weeks 1.0654e-5 months after the MSLB initiation and reactor trip. The RHR cut-in time and the steam release are used as input for the MSLB ANCC radiological dose analysis described in section 3.2.2, MSLB ANCC Dose Analysis (Reference 6.8), of the enclosure to the LAR.

The NRC staff finds that the licensees calculated RHR cut-in time and the steam release rates for the MSLB ANCC condition are acceptable because they are based on the NRC-approved RETRAN02 methodology and conservative inputs and assumptions. The RHR cut-in time after MSLB initiation and steam release are therefore acceptable as inputs for the revised radiological dose analysis.

3.1.3 Evaluation of LRA ANCC Analysis The licensee developed a RETRAN02 model for the LRA analysis in an ANCC situation to determine the extended time to reach RHR cut-in condition and steam releases. The assumptions used in the LRA ANCC analysis, and the comparison between the LRA analysis results obtained by the model to the CLB are described in section 3.3.1, LRA ANCC Cooldown Analysis (Reference 6.2), of the enclosure to the LAR. Table 2 of this SE shows the comparison between the RETRAN02 LRA analysis results with the current analysis of record results.

Table 2: LRA Results Comparison with Current Analysis of Record Scenario with LRA ANCC Current UFSAR Table 15.3-4 RHR cut-in time:

28.0 hrs RHR cut-in time:

8.0 hrs Release from Active SGs (2 SGs):

Release from all SGs (4 SGs):

0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2.360E+05 lbm 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 6.4 E+05 lbm 2 - 28 hours3.240741e-4 days 0.00778 hours 4.62963e-5 weeks 1.0654e-5 months 2.647E+06 lbm 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.120E+06 lbm Release from Inactive SG (2 SGs):

0 - 1.83 hours9.606481e-4 days 0.0231 hours 1.372354e-4 weeks 3.15815e-5 months 2.306E+05 lbm 12.7 - 12.73 hours8.449074e-4 days 0.0203 hours 1.207011e-4 weeks 2.77765e-5 months 3.443E+03 lbm 27.5 - 28 hours3.240741e-4 days 0.00778 hours 4.62963e-5 weeks 1.0654e-5 months 9.620E+03 lbm Total SG Steam Release:

3.127E+06 lbm Total SG Steam Release 1.76E+06 lbm The CLB analysis assumes an 8-hour cooldown to the RHR cut-in condition while the results of the RETRAN02 analysis show that the RHR cut-in conditions are reached in approximately 28 hours3.240741e-4 days 0.00778 hours 4.62963e-5 weeks 1.0654e-5 months after the reactor trip. The RHR cut-in time and steam release are used as input for the LRA ANCC radiological dose analysis described in section 3.3.2, LRA ANCC Dose Analysis (Reference 6.9), of the enclosure to the LAR.

The NRC staff finds that the licensees calculated RHR cut-in time and steam release rates for an LRA ANCC condition are acceptable because they are based on the NRC-approved RETRAN02 methodology and conservative inputs and assumptions. The RHR cut-in time after the LRA initiation and the steam release are therefore acceptable as inputs for the revised radiological dose analysis.

3.1.4 Conformance with RG 1.183 In attachment 1, Regulatory Guide 1.183 Conformance Table, to the LAR, the licensee provided a table which presents the licensees evaluation of the proposed change against the applicable regulatory guidance in RG 1.183. The NRC staff reviewed the evaluations of RG 1.183 related to the thermal-hydraulic analyses in sections 3.6, 5.1.1, 5.1.2, 5.1.3, and 5.2 of this table and determined that the proposed changes conform with the RG 1.183 guidance.

3.1.5 Evaluation of Licensees Comparison of Assumptions and Inputs to MSLB and LRA Thermal-Hydraulic Analysis from their CLB Analysis In attachment 2, Comparison of Assumptions and Inputs to Previous AST Analysis, to the LAR, table 2-1, MSLB Assumptions Comparison, the licensee identified changes in the

assumptions in the proposed MSLB analysis from the CLB analysis and the basis of changes. In table 2-2, MSLB Input Comparison, the licensee identified changes in the input parameters in the proposed analysis from the CLB analysis. In table 2-3, LRA Assumptions Comparison, the licensee identified changes in the assumptions in the proposed LRA analysis from the CLB analysis and the basis of changes. In table 2-4, LRA Input Comparison, the licensee identified changes in the input parameters in the proposed analysis from the CLB analysis. In a letter dated October 19, 2023, the licensee provided justifications of differences in assumptions and the analysis inputs between the proposed analyses and the CLB analysis. The NRC staff reviewed the updated assumptions and inputs and justifications of the differences between the proposed and the CLB thermal-hydraulic MSLB and LRA analyses and finds them acceptable because the changes in the assumption and inputs are reasonable and conservative.

3.1.6 Conclusion for Evaluation of Thermal-Hydraulic Analysis Changes Based on the above technical evaluation of the licensees MSLB and LRA ANCC thermal-hydraulic analyses, the NRC staff determined that the calculated steam release and RHR cut-in time are acceptable as inputs to the dose consequence analysis for these accidents.

There is no impact on the CLB compliance with GDC 10. The licensees compliance with 10 CFR 50.67 is addressed in section 3.2 of this SE.

3.2 Evaluation of Radiological Dose Analysis Changes The NRC staff performed an independent review of inputs, assumptions, and initial conditions in the licensee dose assessment files as necessary to ensure a thorough understanding of the licensees methods, and to verify that values used in the dose assessment analysis were in line with values provided in the LAR, as supplemented.

3.2.1 LRA Evaluation For its reanalysis of the LRA accident, the licensee evaluated an LRA initiated with the instantaneous seizure of a reactor coolant pump (RCP) rotor which causes a rapid reduction in the flow through the affected RCS loop. The sudden decrease in core coolant flow causes a reactor trip on a low flow signal. The low coolant flow causes a degradation of core heat transfer, resulting in localized temperature and pressure changes in the core. The licensees evaluation indicates that the fuel will experience a departure from nucleic boiling, which results in fuel cladding damage. Activity from the fuel cladding damage is transported to the secondary side due to primary-to-secondary side leakage. During an ANCC condition, the primary system cooldown rate must be reduced such that the RCS loop flow through any inactive SG is not stagnated. Therefore, the duration of the post-LRA cooldown must be significantly extended when compared to events that do not account for ANCC. The licensees CLB and current calculations in the application assume a total primary-to-secondary leak rate equal to 1.0 gallons per minute (gpm), which is higher than the current technical specification (TS) total allowable leak rate of 0.42 gpm. It is assumed that the LRA does not cause an increase in the magnitude of the pre-existing primary-to-secondary leakage. Since the radiological dose resulting from an LRA is caused by radioactive material that is released along with steam vented to the environment, an increase in the amount of time that steam is vented results in an increase in the calculated radiological dose to the control room (CR), technical support center (TSC) and to members of the public at the exclusion area boundary (EAB) and low population zone (LPZ).

The LRA accident is described in section 15.3.3.3, Radiological Consequences, of the STP UFSAR. RG 1.183, appendix G, Assumptions for Evaluating the Radiological Consequences of a PWR [Pressurized-Water Reactor] Locked Rotor Accident, identifies acceptable radiological analysis assumptions for a PWR LRA. A summary of the licensees LRA analysis assumptions is included in section 3.3.1 of the enclosure to the LAR.

3.2.1.1 Source Term LAR table 9, LRA Source Term Summary, provides the exact same CLB source term values as provided in table 4.2-20, Initial RCS (@60 µCi/gm [microcuries per gram]) and Secondary Concentrations (@ 0.1 µCi/gm DEI [Dose Equivalent Iodine]) (Noble Gases based on 1% Failed Fuel), of the licensees AST submittal dated March 22, 2007 (Reference 11).

The licensee assumed that the instantaneous seizure of the RCP rotor associated with the LRA results in a small percentage of fuel clad damage. As in the CLB, the dose analysis for this event conservatively assumes 10 percent fuel clad damage with no fuel melt predicted.

Therefore, the source term available for release is associated with this fraction of damaged fuel cladding and the fraction of core activity existing in the gap. The source term is assumed to be released instantaneously and homogeneously through the RCS.

The licensee included iodine in the RCS due to a design-basis pre-accident 60 µCi/gm DEI spike. The licensee included the RCS noble gas activity associated with assumed 1 percent fuel defects. The licensee included the RCS cesium and rubidium concentrations corresponding to 1 percent failed fuel; however, the cesium and rubidium concentrations are not assumed to spike along with the iodines.

The licensee incorporated release fractions and transport fractions that are consistent with RG 1.183, appendix G, and RG 1.183, table 3. In accordance with RG 1.183, 5 percent of the core inventory of iodine and noble gas is assumed to be in the fuel-clad gap, with the exception of 8 percent assumed for Iodine-131 and 10 percent assumed for Krypton 85. Additionally, the licensee included 12 percent of the core cesium and rubidium as shown in table 3 of RG 1.183.

The source term model also includes the maximum TS equilibrium secondary coolant activity concentration of 0.1 Ci/gm DEI. The NRC staff finds that these values are consistent with the AST CLB and are therefore acceptable.

3.2.1.2 Release Transport Consistent with its CLB analysis, with the exception of iodine speciation for releases from SGs, the licensee followed the guidance in RG 1.183, appendix G, assumption 4 in all other aspects of the transport analysis for the LRA. The activity that originates in the RCS is released to the secondary coolant by means of the primary-to-secondary coolant leak rate. The licensee assumed a conservative value for the design-basis leak rate of 1.0 gpm. This 1.0 gpm leak rate is modeled as 0.65 gpm for two SGs with tube coverage and 0.35 gpm for the inactive and faulted SGs with uncovered tubes. For the SG on the loop with the locked RCP rotor, the licensee assumed that the SG tubes become uncovered due to a feedwater isolation valve malfunction. A LOOP is assumed to occur concurrently with the reactor trip, which results in releases to the environment associated with the secondary coolant steaming from the SGs.

Because of the release dynamic of the activity from the SG power operated relief valves (PORVs), RG 1.183 allows for a reduction in the amount of activity released to the environment based on partitioning of nuclides between the liquid and gas states of water for this release

path. For iodine, the partitioning coefficient of 100 was taken directly from RG 1.183. Due to their volatility, 100 percent of the noble gases are assumed to be released. No partitioning is modeled in the inactive SGs. These assumptions are consistent with the CLB values provided with the licensees AST.

The licensees CLB assumes that the steaming release from the PORVs and primary-to-secondary coolant leakage end after 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , at which time the RCS and the secondary system have reached pressure equilibrium. The major change to the CLB is that this application assumes the release continues for 28 hours3.240741e-4 days 0.00778 hours 4.62963e-5 weeks 1.0654e-5 months due to an ANCC, at which point the majority of the release is terminated and the RHR cut-in condition is reached. This change conforms with RG 1.183, appendix G, assumption 5.3 and assumes the release continues for this entire duration until shutdown cooling is in operation and releases from the SGs have been terminated. The total mass released from the SGs to the environment is 3.13 x 106 Ibm over 28 hours3.240741e-4 days 0.00778 hours 4.62963e-5 weeks 1.0654e-5 months . This is an increase from the CLB of 1.76 x 106 Ibm in the CLB. Further, the licensees CLB assumes that minor leakage via the main steam isolation valve (MSIV) above-seat drain orifices continues for 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months , and this is maintained in this calculation.

The licensee used the RADTRAD computer code to model the time dependent transport of radionuclides, from the primary-to-secondary side and consequently to the environment via the PORVs or main steam safety valves. Based on its review, the NRC staff finds that the licensees analysis conforms with appendix G of RG 1.183, which identifies acceptable radiological analysis assumptions for an LRA.

3.2.1.3 CR Habitability for LRA All inputs, assumptions, and initial conditions used for CR habitability remain unchanged from those values used in the licensees AST submittal dated March 22, 2007. The NRC staff has previously evaluated the licensees assumptions for CR dose consequences and found them acceptable as reflected in the CLB (Reference 4).

3.2.1.4 Atmospheric Dispersion Factors All inputs, assumptions, and initial conditions used for atmospheric dispersion factors (/Q) remain unchanged from those values used in the licensees AST submittal dated March 22, 2007, which the NRC staff has previously evaluated and found acceptable as reflected in the CLB (Reference 4).

3.2.1.5 LRA Evaluation Conclusion The licensee evaluated the radiological consequences resulting from the postulated LRA and concluded that the radiological consequences at the EAB, LPZ, and CR are within the radiation dose reference values provided in 10 CFR 50.67 and the accident specific dose criteria specified in SRP Section 15.0.1. The NRC staff finds that the licensee used analysis assumptions and inputs consistent with applicable regulatory guidance identified in section 2.0 of this SE. The assumptions found acceptable to the NRC staff are presented in section 3.3 LRA Dose Consequence Analysis, of the enclosure to the LAR, and the licensees calculated dose results are given in LAR table 11, Locked Rotor Doses for Offsite, Control Room and TSC. The licensees results are given in table 3 of this SE. The NRC staff finds that the EAB, LPZ, and CR radiological doses for the LRA meet the dose reference values provided in 10 CFR 50.67 and the applicable accident dose guidelines and criteria, and, therefore, the licensees analysis is acceptable.

Table 3: Locked Rotor Doses for Offsite, CR and TSC Location TEDE Dose (REM)

Acceptance Criteria (REM)

EAB 1.90 2.5 LPZ 1.66 2.5 CR 4.57 5.0 TSC 4.43 5.0 3.2.2 MSLB Accident Evaluation For its reanalysis of the MSLB accident, the licensee evaluated a MSLB that impacts one of the four SGs. The reanalysis is necessary to update and validate radiological dose assessments that result from an MSLB coupled with a LOOP. The LOOP renders RCPs unavailable and, along with the faulted SG, results in an ANCC condition. During a natural circulation cooldown, flow through the RCS is maintained as a result of a thermal driving head created by differences in temperature and thus density of water across the RCS; therefore, this method of flow is highly impacted by cooldown rates. During an ANCC condition, the primary system cooldown rate must be reduced such that the RCS loop flow through any inactive SG is not stagnated.

Therefore, the duration of the post-MSLB cooldown must be significantly extended when compared to events that do not account for an ANCC. After a MSLB with a concurrent LOOP, the RCS can only be cooled down by steam venting to the environment. The increase in cooldown duration that is necessary to avoid loop stagnation during an ANCC thus increases the amount of time that steam venting is required. Since the radiological dose resulting from an MSLB is caused by radioactive material that is released along with steam vented to the environment, an increase in the amount of time that steam is vented results in an increase in the calculated radiological dose to the CR, TSC and to members of the public at the EAB and LPZ.

The MSLB accident is described in section 15.1.5 of the STP UFSAR. RG 1.183, appendix E, Assumptions for Evaluating the Radiological Consequences of a PWR Main Steam Line Break Accident, identifies acceptable radiological analysis assumptions for a PWR MSLB. A summary of the licensees MSLB analysis assumptions is included in section 3.2.1 of the enclosure to the LAR. The postulated MSLB accident assumes a double-ended break of one main steam line outside the primary containment. The steam release from a rupture of a main steam line would result in initial increase in steam flow, which decreases during the accident as the steam pressure decreases. The increased energy removal from the RCS causes a reduction of coolant temperature and pressure. In the presence of a negative moderator temperature coefficient, the cooldown results in an insertion of positive reactivity. If the most reactive rod cluster control assembly is assumed stuck in its fully withdrawn position after the reactor trip, there is an increased possibility that the core will become critical and continue to generate heat. The core is ultimately shut down by the boric acid delivered by the safety injection system.

In its analysis, the licensee assumed the B SG to be impacted by the MSLB; it is thus referred to as the faulted SG for this SE. The affected SG, SG B, rapidly depressurizes and releases the initial contents of the SG to the environment. The resultant depressurization of the steam system causes the MSIVs to close and, if the plant is operating at power when the event is initiated, causes the reactor to scram. Concurrently, a LOOP occurs with a limiting single failure of ESF train A. The loss of ESF train A eventually results in a loss of auxiliary feedwater to the D SG making it unavailable for steaming; it is thus referred to as inactive for this SE.

Therefore, SGs A and C are used for natural circulation cooldown and are referred to as active SGs for this SE.

The radiological consequences of an MSLB outside containment will bound the consequences of a break inside containment. Therefore, only the MSLB outside containment is considered by the licensee with regard to the radiological consequences. The NRC staff agrees that an MSLB outside containment will be bounding and is sufficient to evaluate with regard to the radiological consequences.

3.2.2.1 Source Term RG 1.183, appendix E, assumption 2, states that if no, or minimal, fuel damage is postulated for the limiting event, the released activity should be the maximum coolant activity allowed by the TSs, including the effects of pre-accident and concurrent iodine spiking. The licensees evaluation indicates that no fuel damage would occur as a result of an MSLB accident.

Therefore, the source term for the radiological release is based on the TS allowed levels of radioactivity in the RCS and the secondary system. The licensee made appropriate assumptions to support its analysis: (1) an initial (i.e., preceding the accident) SG radioisotope inventory consistent with the equilibrium secondary-side specific activity TS limiting condition for operation of 0.1 µCi/gm DEI; (2) a noble gas concentration consistent with the 1 percent failed fuel assumption; and (3) the two RCS radioiodine spiking cases described in RG 1.183.

The first RCS radioiodine case is referred to as a pre-incident iodine spike and assumes that a reactor transient has occurred prior to the postulated MSLB that has raised the primary coolant iodine concentration to the maximum value permitted by the TS for a spiking condition. For STP, the maximum iodine concentration allowed by the TS as the result of an iodine spike is 60 µCi/gm DEI.

The second RCS radioiodine case assumes that the primary system transient associated with the MSLB causes an iodine spike in the primary system. The case is referred to as a coincident iodine spike. The increase in primary coolant iodine concentration for the coincident iodine spike case is estimated using a spiking model that assumes that the iodine release rate from the fuel rods to the primary coolant increases to a value 500 times greater than the release rate corresponding to the iodine concentration at the TS limit for normal operation. For STP, the RCS TS limit for normal operation is 1.0 µCi/gm. The duration of the coincident iodine spike is assumed to be 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months in accordance with the applicable guidance. Additionally, since RG 1.183, specifies that the chemical form of particulate iodine is cesium iodide (CsI), the licensee assumed that the primary system transient that causes the iodine spiking also increases the cesium and rubidium concentrations in the RCS in relative amounts.

RG 1.183, appendix E, assumption 4 states that, The chemical form of radioiodine released from the fuel should be assumed to be 95% cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide. Iodine releases from the steam generators to the environment should be assumed to be 97% elemental and 3% organic. These fractions apply to iodine released as a result of fuel damage and to iodine released during normal operations, including iodine spiking.

In its CLB, the licensee took a conservative approach to the RG 1.183, appendix E statement that iodine release from the SG should be 97 percent elemental and 3 percent organic. In this

LAR, consistent with its CLB analysis, the licensee assumed, for the active SGs, a partition factor of 100 for particulate and elemental iodine and a partition factor of 1 for organic iodine.

This results in an effective radioiodine mix of 4.2 percent elemental, 13.1 percent organic, and 82.7 percent particulate. Since the proposed source term is consistent with or more conservative than the RG 1.183 guidance, the NRC staff finds the source term acceptable.

3.2.2.2 Transport Consistent with its CLB analysis, with the exception of iodine speciation for releases from SGs discussed in section 3.2.2.1 of this SE, the licensee followed the guidance in RG 1.183, appendix E, assumption 5 in all other aspects of the transport analysis for the MSLB.

For additional conservatism, the licensee assumed a total primary-to-secondary leak (PTSL) rate equal to 1.0 gpm which is higher than the TS total allowable leak rate of 0.42 gpm. The licensee modeled the assumed 1.0 gpm PTSL as 0.35 gpm to the inactive and faulted SGs (SGs A and D) and 0.65 gpm to the active SGs (SGs A and C). The 0.35 gpm portion of the PTSL is modeled as a release direct to the environment from the RCS.

RG 1.183, appendix E, assumption 5.2, states that, The density used in converting volumetric leak rates (e.g., gpm) to mass leak rates (e.g., [pounds of mass per hour] lbm/hr) should be consistent with the basis of the parameter being converted. The ARC [alternate repair criteria] leak rate correlations are generally based on the collection of cooled liquid. Surveillance tests and facility instrumentation used to show compliance with leak rate technical specifications are typically based on cooled liquid. In most cases, the density should be assumed to be 1.0 gm/cc [grams per cubic centimeter]

(62.4 [pounds of mass per cubic foot] lbm/ft3).

The licensees leak rate testing results are adjusted so that the allowable leakage corresponds to a density of 8.33 pounds per gallon and, consistent with the CLB, this density was used to convert the volumetric leak rate to a total mass flow rate due to SG tube leakage in the MSLB dose consequence analysis.

RG 1.183, appendix E, assumption 5.3, states that, The primary-to-secondary leakage should be assumed to continue until the primary system pressure is less than the secondary system pressure, or until the temperature of the leakage is less than 100oC [degrees Celsius] (212 oF). The release of radioactivity from unaffected steam generators should be assumed to continue until shutdown cooling is in operation and releases from the steam generators have been terminated.

In accordance with RG 1.183, and the CLB, the license assumed that PTSL continues for 28 hours3.240741e-4 days 0.00778 hours 4.62963e-5 weeks 1.0654e-5 months following the MSLB after which time the RHR cut-in condition is reached.

In accordance with RG 1.183, the licensee assumed that all noble gas radionuclides released from the primary system are released to the environment without reduction or mitigation. The licensee modeled the RCS noble gas source term as a direct release path from the primary coolant to the environment. Additionally, for conservatism, the licensee included noble gases in the source terms that are assumed to be released to and through the SG volumes, as well.

Although not necessary for consistency with RG 1.183, the NRC staff finds the addition of noble gases to the SG volume source terms acceptable because this results in a more conservative case.

RG 1.183, appendix E, assumption 5.5.4, states that, The radioactivity in the bulk water is assumed to become vapor at a rate that is the function of the steaming rate and the partition coefficient. A partition coefficient for iodine of 100 may be assumed. The retention of particulate radionuclides in the steam generators is limited by moisture carryover from the steam generators.

Accordingly, the licensee assumed that the radioactivity in the bulk water of the active SGs becomes vapor at a rate that is a function of the steaming rate and the partition coefficient. The licensee used a partition coefficient of 100 for elemental iodine and other particulate radionuclides released from the intact SGs. As discussed above, the licensee did not credit the partitioning of organic iodine. The licensee assumed that the organic iodine migrates directly to the steam space and becomes immediately available for release.

The accident includes an MSLB that faults SG B coincident with a LOOP with the single failure of ESF train A, resulting in a loss of auxiliary feed water (AFW) flow to SGs A and D and leaving SG C unaffected. The licensee assumed that AFW is cross-connected at 30 minutes after the MSLB initiation to provide feedwater to SG A. Therefore, SGs A and C are used for natural circulation cooldown. Because SG D has no AFW flow, it dries out and is modeled to appropriately partition its iodine contribution. Additionally, at 30 minutes, AFW flow to the faulted SG B is isolated by operator action.

The licensees analysis separates steam releases from faulted SG B, inactive SG D and active SGs A and C. For discretionary margin for any reanalysis of steam releases, the steam releases from the licensees thermodynamic analysis have been increased 20 percent and some of the faulted and inactive SG steam release intervals are extended. Section 2.0 of STP Calculation NC07143, MSLB Steam Release for Dose Analysis, Revision 2, included in attachment 4, Dose Analysis Calculation and Code Output Files, to the LAR, provides the results of the thermodynamic analysis of the event. In general, the impact of considering an ANCC condition involves a significant reduction in assumed cooldown rate and thus an extension in the duration of releases from the faulted SG to the environment. The steam release to the environment through the relief valves is assumed to last for 28 hours3.240741e-4 days 0.00778 hours 4.62963e-5 weeks 1.0654e-5 months at which point the RHR cut-in condition is met. Consistent with the CLB, the steam release through the above seat valve is conservatively assumed to last for 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

Consistent with the CLB, the licensee assumed that the break and the above seat drain releases occur in the isolation valve cubicle next to the PORVs. Therefore, the PORV-to-CR

/Qs are used for the CR and TSC dose analyses. Since the proposed transport analysis is consistent with or more conservative than the RG 1.183 guidance, the NRC staff finds the transport analysis acceptable.

3.2.2.3 CR Ventilation Assumptions for the MSLB The licensee evaluated CR habitability for the MSLB consistent with approaches described in its CLB and RG 1.183, section 4.2, Control Room Dose Consequences. The NRC staff has

previously evaluated the licensees assumptions for CR dose consequences and found them acceptable as reflected in the CLB (Reference 4).

3.2.2.4 Atmospheric Dispersion Factors All inputs, assumptions, and initial conditions used for /Q remain unchanged from those values used in the licensees AST submittal dated March 22, 2007, which the NRC staff has previously evaluated and found acceptable as reflected in the CLB (Reference 4).

3.2.2.5 MSLB Accident Evaluation Conclusion The licensee evaluated the radiological consequences resulting from the postulated MSLB incident and concluded that the radiological consequences at the EAB, LPZ and CR are within the dose reference values provided in 10 CFR 50.67 and the accident specific dose criteria specified in SRP Section 15.0.1 as well as in table 6 of RG 1.183. The NRC staff finds that the licensee used analysis assumptions and inputs consistent with applicable regulatory guidance identified in section 2.0 of this SE. The assumptions found acceptable to the NRC staff are presented in section 3.2 MSLB Dose Consequence Analysis, of the enclosure to the LAR, and the licensees calculated dose results are given in LAR table 7, MSLB Doses for Offsite, Control Room and TSC. The licensees results are given in table 4 of this SE. Based on its review, the staff finds that the EAB, LPZ and CR doses estimated by the licensee for the MSLB meet the dose reference values provided in 10 CFR 50.67, and the applicable accident dose guidelines and criteria, and, therefore, the licensees analysis is acceptable.

Table 4: MSLB Doses for Offsite, CR and TSC Coincident Iodine Spike Pre-accident Iodine Spike Result (rem TEDE)

Limit (rem TEDE)

Result (rem TEDE)

Limits (rem TEDE)

EAB (Worst two-hours) 1.30 2.5 0.097 25 LPZ (0-30 days) 1.16 2.5 0.081 25 CR (0-30 days) 3.39 5

0.237 5

TSC (0-30 days) 3.30 5

0.231 5

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Texas State official was notified of the proposed issuance of the amendments on January 19, 2024. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve 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 previously issued a proposed finding that the amendments involve no significant hazards consideration published in the Federal Register on June 13, 2023 (88 FR 38552), and there has

been no public comment on such finding. Accordingly, the amendments meet 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 amendments.

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) there is reasonable assurance that 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.

Harshaw, K. A., STPNOC, letter to NRC, License Amendment Request to Revise Alternative Source Term Dose Calculation, dated March 30, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23089A204).

2.

Harshaw, K. A., STPNOC, letter to NRC, Response to Request for Additional Information for License Amendment Request to Revise Alternate Source Term Dose Calculation, dated October 19, 2023 (ML23292A311).

3.

Harshaw, K. A., STPNOC, letter to NRC, Supplement to Modify the lmplementation Date for License Amendment Request to Revise Alternate Source Term Dose Calculation (EPID: L-2023-LLA-0047), dated February 8, 2024 (ML24039A160).

4.

Thadani, M. C., NRC, letter to J. J. Sheppard, STPNOC, South Texas Project, Units 1 and 2 - Issuance of Amendment Nos. 182 and 169, Regarding Adoption of Alternate Radiological Source Term in Assessment of Design-Basis Accident Dose Consequences (TAC Nos. MD4996 and MD4997), dated March 6, 2008 (Package ML080300062).

5.

Wink, R., Ameren Missouri, letter to NRC, Docket Number 50-483, Callaway Plant, Unit 1, Union Electric Co. Renewed Facility Operating License NPF-30, Licensee Event Report 2018-002-00, Inadequate EOP guidance for Asymmetric Natural Circulation Cooldown, dated July 3, 2018 (ML18184A389).

6.

NRC, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Reactors, RG 1.183, Revision 0, dated July 2000 (ML003716792).

7.

NRC, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, NUREG-0800, Section 15.0.1, Revision 0, Radiological Consequence Analyses Using Alternative Source Terms, dated July 2000, (ML003734190).

8.

Westinghouse, RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses, WCAP-14882-P-A, dated April 1999 (ML093421329; not publicly available, proprietary information).

9.

Information Systems Laboratory, Inc. and NRC, SNAP/RADTRAD 4.0: Description of Models and Methods, NUREG/CR-7220, dated June 2016 (ML16160A019).

10.

STPNOC, South Texas Project, Units 1 and 2, Updated Final Safety Analysis Report Revision 21, Chapter 15, Accident Analyses, dated April 28, 2022 (ML22140A057).

11.

Rencurrel, D. W., STPNOC, South Texas Project Units 1 and 2, Docket Nos. STN50-498, STN 50-499, Request for License Amendment Related to Application of the Alternate Source Term, dated March 22, 2007 (ML070890474).

Principal Contributors:

Noushin Amini David Garmon Sean Meighan Ahsan Sallman Date : February 20, 2024

ML24022A225

concurrence by email OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA*

NRR/DSS/SNSB/BC*

NAME DGalvin PBlechman PSahd DATE 1/24/2024 1/24/2024 11/8/2023 OFFICE NRR/DRA/ARCB/BC*

OGC / NLO*

NRR/DORL/LPL4/BC*

NAME KHsueh AGhosh Naber JRankin DATE 1/12/2024 2/15/2024 2/20/2024 OFFICE NRR/DORL/LPL4/PM*

NAME DGalvin DATE 2/20/2024

v t e Letter Sequence Approval
EPID:L-2023-LLA-0047, License Amendment Request to Revise Alternative Source Term Dose Calculation (Approved, Closed)
Initiation Request Acceptance Supplement Administration Questions 12 September 2023 21 September 2023 Answers Results Approval Other:
MONTHYEAR2023-09-12 [Table View]