Text

Issuance of Amendment Nos. 305 and 195 to Modify Primary and Secondary Coolant Activity Technical Specifications
	ML20213A731
Person / Time
Site:	Beaver Valley
Issue date:	09/23/2020
From:	Jennifer Tobin; Plant Licensing Branch 1
To:	Penfield R; Energy Harbor Nuclear Corp
	Tobin J
References
	Download: ML20213A731 (64)
	v • d • e

September 23, 2020 Mr. Rod L. Penfield Energy Harbor Nuclear Corp.

Beaver Valley Power Station Mail Stop P-BV-SSB P.O. Box 4, Route 168 Shippingport, PA 15077-0004

SUBJECT:

BEAVER VALLEY POWER STATION, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 305 AND 195 TO MODIFY PRIMARY AND SECONDARY COOLANT ACTIVITY TECHNICAL SPECIFICATIONS (EPID L-2019-LLA-0223)

Dear Mr. Penfield:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment Nos. 305 and 195 to Renewed Facility Operating License Nos. DPR-66 and NPF-73, respectively, for the Beaver Valley Power Station, Units 1 and 2. These amendments consist of changes to the technical specifications in response to your application dated October 20, 2019, as supplemented by letters dated April 14, 2020, and June 9, 2020.

The amendments revise Technical Specifications 3.4.16, RCS [Reactor Coolant System]

Specific Activity; 3.7.13, Secondary Specific Activity; 5.5.7, Ventilation Filter Testing Program (VFTP); and 5.5.14, Control Room Envelope Habitability Program. The amendments modify the primary and secondary coolant activities and control room emergency ventilation system testing criteria and permit a one-time change to the common control room envelope unfiltered air inleakage test frequency.

R. Penfield 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/

Jennifer C. Tobin, Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-334 and 50-412

Enclosures:

1. Amendment No. 305 to DPR-66

2. Amendment No. 195 to NPF-73

3. Safety Evaluation cc: Listserv

ENERGY HARBOR NUCLEAR CORP.

ENERGY HARBOR NUCLEAR GENERATION LLC DOCKET NO. 50-334 BEAVER VALLEY POWER STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 305 Renewed License No. DPR-66

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

A. The application to amend Renewed Facility Operating License No. DPR-66 filed by FirstEnergy Nuclear Operating Company (FENOC),*, ** acting on its own behalf and as agent for FirstEnergy Nuclear Generation, LLC*** (the licensees),

dated October 20, 2019, as supplemented by letters dated April 14, 2020, and June 9, 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, 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.

FirstEnergy Nuclear Operating Company has been renamed Energy Harbor Nuclear Corp.

- Energy Harbor Nuclear Corp. is authorized to act as agent for Energy Harbor Nuclear Generation LLC and has exclusive responsibility and control over the physical construction, operation, and maintenance of the facility.

- - FirstEnergy Nuclear Generation, LLC has been renamed Energy Harbor Nuclear Generation LLC.

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. DPR-66 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 305, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by James James G. G. Danna Date: 2020.09.23 Danna 12:12:05 -04'00' James G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: September 23, 2020

ATTACHMENT TO LICENSE AMENDMENT NO. 305 BEAVER VALLEY POWER STATION, UNIT NO. 1 RENEWED FACILITY OPERATING LICENSE NO. DPR-66 DOCKET NOS. 50-334 Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

Remove Insert Page 3 Page 3 Replace the following pages of the 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.

Remove Insert 3.4.16-1 3.4.16-1 3.4.16-2 3.4.16-2 3.4.16-3 3.4.16-3

--- 3.4.16-4 3.7.13-1 3.7.13-1 5.5-15 5.5-15 5.5-21 5.5-21 5.5-22 5.5-22

(3) Energy Harbor Nuclear Corp., pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Energy Harbor Nuclear Corp., pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (5) Energy Harbor Nuclear Corp., pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter 1:

Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level Energy Harbor Nuclear Corp. is authorized to operate the facility at a steady state reactor core power level of 2900 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 305, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

(3) Auxiliary River Water System (Deleted by Amendment No. 8)

Amendment No. 305 Beaver Valley Unit 1 Renewed Operating License DPR-66

RCS Specific Activity 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity LCO 3.4.16 The specific activity of the reactor coolant shall be within limits.

APPLICABILITY: MODES 1 and 2, MODE 3 with RCS average temperature (Tavg) 500F.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. DOSE EQUIVALENT ------------------------------------------------

I-131 > 0.35 Ci/gm - NOTE -

(Unit 1), and > 0.10 Ci/gm LCO 3.0.4.c is applicable.

(Unit 2). ------------------------------------------------

A.1 Verify DOSE Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months EQUIVALENT I-131 within the acceptable region of Figure 3.4.16-1 (Unit 1),

and Figure 3.4.16-2 (Unit 2).

AND A.2 Restore DOSE 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months EQUIVALENT I-131 to within limit.

B. Gross specific activity of B.1 Be in MODE 3 with 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months the reactor coolant not Tavg < 500F.

within limit.

C. Required Action and C.1 Be in MODE 3 with 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Tavg < 500F.

Time of Condition A not met.

OR DOSE EQUIVALENT I-131 in the unacceptable region of Figure 3.4.16-1 (Unit 1),

and Figure 3.4.16-2 (Unit 2).

Beaver Valley Units 1 and 2 3.4.16 - 1 Amendments 305 / 195

RCS Specific Activity 3.4.16 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 Verify reactor coolant gross specific activity In accordance 100/ E Ci/gm. with the Surveillance Frequency Control Program SR 3.4.16.2 ------------------------------------------------------------------------

- NOTE -

Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT I-131 In accordance specific activity 0.35 Ci/gm (Unit 1), and with the 0.10 Ci/gm (Unit 2). Surveillance Frequency Control Program AND Between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a THERMAL POWER change of 15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period SR 3.4.16.3 ------------------------------------------------------------------------

- NOTE -

Not required to be performed until 31 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

Determine E from a sample taken in MODE 1 after a In accordance minimum of 2 effective full power days and 20 days of with the MODE 1 operation have elapsed since the reactor was Surveillance last subcritical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . Frequency Control Program Beaver Valley Units 1 and 2 3.4.16 - 2 Amendments 305 / 195

RCS Specific Activity 3.4.16 REACTOR COOLANT DOSE EQUIVALENT I-131 SPECIFIC ACTIVITY LIMIT ( Ci/gm) 100 Figure 3.4.16-1 (Page 1 of 1)

Reactor Coolant DOSE EQUIVALENT I-131 Specific Activity Limit Versus Percent of RATED THERMAL POWER (Unit 1)

Beaver Valley Units 1 and 2 3.4.16 - 3 Amendments 305 / 195

RCS Specific Activity 3.4.16 REACTOR COOLANT DOSE EQUIVALENT I-131 SPECIFIC ACTIVITY LIMIT ( Ci/gm)

Figure 3.4.16-2 (Page 1 of 1)

Reactor Coolant DOSE EQUIVALENT I-131 Specific Activity Limit Versus Percent of RATED THERMAL POWER (Unit 2)

Beaver Valley Units 1 and 2 3.4.16 - 4 Amendments 305 / 195

Secondary Specific Activity 3.7.13 3.7 PLANT SYSTEMS 3.7.13 Secondary Specific Activity LCO 3.7.13 The specific activity of the secondary coolant shall be 0.10 Ci/gm DOSE EQUIVALENT I-131 (Unit 1), and 0.05 Ci/gm DOSE EQUIVALENT I-131 (Unit 2).

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Specific activity not within A.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months limit.

AND A.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.13.1 Verify the specific activity of the secondary coolant is In accordance 0.10 Ci/gm DOSE EQUIVALENT I-131 (Unit 1), and with the 0.05 Ci/gm DOSE EQUIVALENT I-131 (Unit 2). Surveillance Frequency Control Program Beaver Valley Units 1 and 2 3.7.13 - 1 Amendments 305 / 195

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testing Program (VFTP) (continued)

ESF Ventilation System Penetration Flowrate SLCRS 1.0% (Unit 1) 32,400 cfm and 39,600 cfm (Unit 1) 0.05% (Unit 2) 51,300 cfm and 62,700 cfm (Unit 2)

CREVS 0.5% 800 cfm and 1000 cfm

c. Demonstrate for each of the required ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, or using a slotted tube sampler in accordance with ANSI N509-1980 shows, within 31 days after removal, the methyl iodide removal efficiency greater than or equal to the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of 30C, an inlet methyl iodide concentration of 1.75 mg/m3, and an air flow velocity and relative humidity (RH) specified below:

ESF Ventilation Removal System Efficiency Air Flow Velocity RH SLCRS 90% (Unit 1) 0.9 ft/sec (Unit 1) 95% (Unit 1) 99% (Unit 2) 0.7 ft/sec (Unit 2) 70% (Unit 2)

CREVS 99.5% (Unit 1) 0.68 ft/sec (Unit 1) 70% (Unit 1) 99.5% (Unit 2) 0.7 ft/sec (Unit 2) 70% (Unit 2)

d. Demonstrate for each of the required ESF systems that the pressure drop across the combined HEPA filters and the charcoal adsorbers is less than the value specified below when tested at the system flowrate specified as follows:

ESF Ventilation System Delta P Flowrate SLCRS 6 inches Water Gauge 32,400 cfm and 39,600 cfm (Unit 1) (Unit 1) 6.8 inches Water 51,300 cfm and 62,700 cfm Gauge (Unit 2) (Unit 2)

CREVS 6 inches Water Gauge 800 cfm and 1000 cfm (Unit 1)

(Unit 1) 5.6 inches Water 800 cfm and 1000 cfm (Unit 2)

Gauge (Unit 2)

Beaver Valley Units 1 and 2 5.5 - 15 Amendments 305 / 195

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.14 Control Room Envelope Habitability Program (continued)

c. -------------------------------------------------------------------------------------------------

- NOTE -

The three-year test frequency for the CRE unfiltered air inleakage test failure that occurred in October 2017 may be extended an additional three years, not to exceed October 2023.

Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CREVS, operating at the flow rate required by the VFTP, at a Frequency of 18 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the periodic assessment of the CRE boundary.

e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c.

The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences.

Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f. The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.

5.5.15 Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.

Beaver Valley Units 1 and 2 5.5 - 21 Amendments 305 / 195

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.15 Surveillance Frequency Control Program (continued)

b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies, Revision 1.

c. The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.

Beaver Valley Units 1 and 2 5.5 - 22 Amendments 305 / 195

ENERGY HARBOR NUCLEAR CORP.

ENERGY HARBOR NUCLEAR GENERATION LLC DOCKET NO. 50-412 BEAVER VALLEY POWER STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 195 Renewed License No. NPF-73

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

A. The application to amend Renewed Facility Operating License No. NPF-73 filed by FirstEnergy Nuclear Operating Company (FENOC),*, ** acting on its own behalf and as agent for FirstEnergy Nuclear Generation, LLC*** (the licensees),

dated October 20, 2019, as supplemented by letters dated April 14, 2020, and June 9, 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, 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.

FirstEnergy Nuclear Operating Company has been renamed Energy Harbor Nuclear Corp.

- Energy Harbor Nuclear Corp. is authorized to act as agent for Energy Harbor Nuclear Generation LLC and has exclusive responsibility and control over the physical construction, operation, and maintenance of the facility.

- - FirstEnergy Nuclear Generation, LLC has been renamed Energy Harbor Nuclear Generation LLC.

Enclosure 2

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-73 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 195, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto and hereby incorporated in the license. Energy Harbor Nuclear Corp. shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of its date of issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by James James G. G. Danna Date: 2020.09.23 Danna 12:12:47 -04'00' James G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: September 23, 2020

ATTACHMENT TO LICENSE AMENDMENT NO. 195 BEAVER VALLEY POWER STATION, UNIT 2 RENEWED FACILITY OPERATING LICENSE NO. NPF-73 DOCKET NO. 50-412 Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

Remove Insert Page 4 Page 4 Beaver Valley Power Station, Unit 2, uses the same Appendix A as Beaver Valley Power Station, Unit 1. Accordingly, the Unit No. 1 Renewed Facility Operating License has been updated with the following pages, which are applicable to both Units 1 and 2:

Remove Insert 3.4.16-1 3.4.16-1 3.4.16-2 3.4.16-2 3.4.16-3 3.4.16-3

--- 3.4.16-4 3.7.13-1 3.7.13-1 5.5-15 5.5-15 5.5-21 5.5-21 5.5-22 5.5-22

C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations set forth in 10 CFR Chapter 1 and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level Energy Harbor Nuclear Corp. is authorized to operate the facility at a steady state reactor core power level of 2900 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 195, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated in the license. Energy Harbor Nuclear Corp. shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Amendment No. 195 Beaver Valley Unit 2 Renewed Operating License NPF-73

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 305 AND 195 TO RENEWED FACILITY OPERATING LICENSE NOS. DPR-66 AND NPF-73 ENERGY HARBOR NUCLEAR CORP.

ENERGY HARBOR NUCLEAR GENERATION LLC BEAVER VALLEY POWER STATION, UNITS 1 AND 2 DOCKET NOS. 50-334 AND 50-412

1.0 INTRODUCTION

By letter dated October 20, 2019 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19293A367), as supplemented by letters dated April 14, 2020, and June 9, 2020 (ADAMS Accession Nos. ML20105A347 and ML20161A206, respectively), Energy Harbor Nuclear Corp. (the licensee) submitted a license amendment request to revise the Beaver Valley Power Station (Beaver Valley), Units 1 and 2, Technical Specifications (TSs).

The proposed changes would revise TS 3.4.16, RCS [Reactor Coolant System] Specific Activity; 3.7.13, Secondary Specific Activity; 5.5.7, Ventilation Filter Testing Program (VFTP); and 5.5.14, Control Room Envelope Habitability Program.

The proposed changes would modify the primary and secondary coolant activities and control room emergency ventilation system (CREVS) testing criteria and permit a one-time change to the common control room envelope (CRE) unfiltered air inleakage test frequency.

The supplemental letters dated April 14, 2020, and June 9, 2020, provided additional information that clarified the license amendment application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC or the Commission) staffs original proposed no significant hazards consideration determination as published in the Federal Register on December 17, 2019 (84 FR 68953).

2.0 REGULATORY REQUIREMENTS Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36(b) requires each license authorizing operation of a production or utilization facility of a type described in Sections 50.21 or 50.22 to include TSs. The TSs are to be derived from the analyses and evaluations included in the safety analysis report and amendments thereto submitted pursuant to 10 CFR 50.34.

Section 50.36(c)(2) of 10 CFR states that limiting conditions for operation (LCOs) are the lowest functional capability or performance levels of equipment required for safe operation of the Enclosure 3

facility. It further states that when an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the condition can be met.

Section 50.36(c)(3) of 10 CFR states that surveillance requirements (SRs) 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 LCOs will be met.

Section 50.36(c)(5) of 10 CFR states that administrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.

2.1 REGULATORY EVALUATION

2.1.1 System Description Accidents or transients may involve secondary steam release to the atmosphere such as a main steam line break (MSLB), a locked rotor accident (LRA), a loss of alternating current power (LACP), a control rod ejection accident (CREA), and to a lesser extent, a steam generator tube rupture (SGTR). The primary-to-secondary leakage contaminates the secondary fluid. The limit on the primary-to-secondary leakage ensures that the dose contribution at the site boundary from tube leakage following such accidents is limited to appropriate fractions of the 10 CFR 50.67(b) limit of 25 roentgen equivalent man (rem) total effective dose equivalent (TEDE), as discussed in Regulatory Guide (RG) 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, dated July 2000 (ADAMS Accession No. ML003716792). The limit on the primary-to-secondary leakage also ensures that the dose contribution from tube leakage in the control room (CR) is limited to the 10 CFR 50.67(b)(2)(iii) limit of 5 rem TEDE for the duration of the accident.

In TS 3.4.16, the specific iodine activity is limited to 0.35 microcuries per gram (Ci/gm) dose equivalent (DE) iodine I-131 (DE I-131 or DEI) (Unit 1) and 0.10 Ci/gm DE I-131 (Unit 2). The gross specific activity in the reactor coolant is limited to the number of Ci/gm equal to 100 divided by E (average disintegration energy of the sum of the average beta and gamma energies of the non-iodine coolant nuclides). The limit on DE I-131 ensures the TEDE at the site boundary and in the CR during the design-basis accident (DBA) will be an appropriate fraction of the allowed TEDE dose. The limit on gross specific activity provides an additional indication of radionuclides (excluding iodines) that corresponds closely to the noble gas activity in the RCS and helps to ensure the effective doses during the DBA will be an appropriate fraction of the allowed dose.

Activity in the secondary coolant results from steam generator (SG) tube outleakage from the reactor coolant system (RCS). Under steady state conditions, the activity is primarily iodines with relatively short half-lives, and thus, indicates current conditions. During transients, I-131 spikes have been observed, as well as increased releases of some noble gases. Other fission product isotopes, as well as activated corrosion products in lesser amounts, may also be found in the secondary coolant. A limit on secondary coolant specific activity during power operation minimizes releases to the environment because of normal operation, anticipated operational occurrences, and accidents. The requirements in TSs 5.5.7 and 5.5.14 are meant to ensure that ventilation systems reduce the activity released to the environment and that operators can safely operate the plant during design-basis events.

RG 1.196 defines the CRE as follows:

Control Room Envelope (CRE): The plant area, defined in the facilitys licensing basis, that in the event of an emergency, can be isolated from the plant areas and the environment external to the CRE. This area is served by an emergency ventilation system, with the intent of maintaining the habitability of the control room. This area encompasses the control room, and may encompass other non-critical areas to which frequent personnel access or continuous occupancy is not necessary in the event of an accident.

TS 5.5.14 was added to the Units 1 and 2 TSs by Amendment Nos. 281 and 163, respectively, dated February 15, 2008 (ADAMS Accession No. ML080370172). The purpose of adding TS 5.5.14 was to assure the habitability of the CRE between performances of unfiltered air inleakage tests on the CRE boundary. This TS is meant to ensure CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The test frequency specified in TS 5.5.14.c provides assurance that significant degradation of the CRE boundary will not go undetected between CRE inleakage determinations.

The NRC staff evaluated the radiological consequences of affected DBAs for Units 1 and 2 as proposed by the licensee against the dose criteria specified in 10 CFR 50.67(b)(2). These criteria are 25 rem TEDE at the exclusion area boundary (EAB) for any 2-hour period following the onset of the postulated fission product release, 25 rem TEDE at the outer boundary of the low population zone (LPZ) for the duration of the postulated fission product release, and 5 rem TEDE for access to and occupancy of the CR for the duration of the postulated fission product release.

This safety evaluation (SE) addresses the impact of the proposed changes on previously analyzed DBA radiological consequences and the acceptability of the revised analysis results.

The regulatory requirements on which the NRC staff based its review are the accident dose criteria in 10 CFR 50.67, as supplemented in Regulatory Position 4.4 of RG 1.183 and NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition (Standard Review Plan or SRP), Rev. 0, Section 15.0.1. The licensee has not proposed any significant deviation or departure from the guidance provided in RG 1.183. The NRC staffs evaluation is based upon the following regulations, regulatory guides, and standards:

10 CFR Part 50.67, Accident source term Appendix A, General Design Criterion for Nuclear Power Plants, to 10 CFR Part 50 (GDC), Criterion 19, Control room RG 1.23, Revision 0, Onsite Meteorological Programs, dated February 17, 1972 RG 1.52, Revision 3, Design, Inspection, and Testing Criteria for Air Filtration and Adsorption Units of Post-Accident Engineered-Safety-Feature Atmosphere Cleanup Systems in Light-Water-Cooled Nuclear Power Plants, dated June 2001 RG 1.145, Revision 1, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants, dated November 1982

RG 1.183, Revision 0, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, dated July 2000 RG 1.194, Revision 0, Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants, dated June 2003 RG 1.196, Revision 0, Control Room Habitability at Light-Water Nuclear Power Reactors, dated May 2003 NUREG-0409, Iodine Behavior in a PWR Cooling System Following a Postulated Steam Generator Tube Rupture Accident, dated May 1985 NUREG-0800, Standard Review Plan, Section 2.3.4, Short-Term Atmospheric Dispersion Estimates for Accident Releases, Revision 3, dated March 2007 NUREG-0800, Standard Review Plan, Section 6.4, Control Room Habitability System, Revision 3, dated March 2007 NUREG-0800, Standard Review Plan, Section 6.5.2, Containment Spray as a Fission Product Cleanup System, Revision 4, dated March 2007 NUREG-0800, Standard Review Plan, Section 15.0.1, Radiological Consequence Analyses Using Alternative Source Terms, Revision 0, dated July 2000 NUREG/CR-5950, Iodine Evolution and pH Control, dated December 1992 NUREG-1465, Accident Source Terms for Light-Water Nuclear Power Plants, dated February 1995 NUREG-0737, Clarification of TMI [Three Mile Island] Action Plan Requirements, dated November 1980 NRC Draft Regulatory Guide DG-1199, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, dated October 2009 (DG-1199) 2.1.2 Description of Technical Specification Changes The TSs for Units 1 and 2 are maintained together in Renewed Facility Operating License No. DPR-66 and are on the same page with the amendment numbers listed for both units. The proposed changes to TSs 3.4.16 and 3.7.13 would add information to specify the unit-specific requirements for the allowed RCS and secondary coolant specific activities at Units 1 and 2.

The actions table of TS 3.4.16 would be changed as indicated below by the addition of the bold, underlined text:

CONDITION REQUIRED ACTION COMPLETION TIME A. DOSE EQUIVALENT -------------------------------------------------

I-131 > 0.35 Ci/gm - NOTE -

(Unit 1), and > 0.10 LCO 3.0.4.c is applicable.

Ci/gm (Unit 2). ------------------------------------------------

A.1 Verify DOSE Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months EQUIVALENT I-131 within the acceptable region of Figure 3.4.16-1 (Unit 1),

and Figure 3.4.16-2 (Unit 2).

AND 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months A.2 Restore DOSE EQUIVALENT I-131 to within limit.

B. Gross specific activity Be in MODE 3 with Tavg < 500F. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months of the reactor coolant not within limit.

C. Required Action and C.1 Be in MODE 3 with Tavg 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion < 500F.

Time of Condition A not met.

OR DOSE EQUIVALENT I-131 in the unacceptable region of Figure 3.4.16-1 (Unit 1), and Figure 3.4.16-2 (Unit 2).

SR 3.4.16.2 would be changed as indicated below by the addition of the bold, underlined text:

SR 3.4.16.2

- NOTE -

Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT I-131 specific activity 0.35 Ci/gm (Unit 1), and 0.10 Ci/gm (Unit 2).

Figure 3.4.16-1 would be annotated with a (Unit 1) at the end of the title, and a new unit-specific figure for reactor coolant DE I-131 specific activity limit versus percent of rated thermal power would be added as Figure 3.4.16-2.

The LCO statement for TS 3.7.13 would be changed as indicated below by the addition of the bold, underlined text:

LCO 3.7.13 The specific activity of the secondary coolant shall be 0.10 Ci/gm DOSE EQUIVALENT I-131 (Unit 1), and 0.05 Ci/gm DOSE EQUIVALENT I-131 (Unit 2).

SR 3.7.13.1 would be changed as indicated below by the addition of the bold, underlined text:

SR 3.7.13.1 Verify the specific activity of the secondary coolant is 0.10 Ci/gm DOSE EQUIVALENT I-131 (Unit 1), and 0.05 Ci/gm DOSE EQUIVALENT I-131 (Unit 2).

The proposed revisions to TS 5.5.7 for the CREVS change the acceptance criteria for the CREVS penetration and system bypass requirement and CREVS charcoal adsorber removal efficiency from 99% to 99.5% for both units based on the licensees evaluation in the license amendment request (LAR).

The proposed change to the CRE habitability program in TS 5.5.14 in the Administrative Controls Section of the TSs would add a note allowing a one-time extension of 3 years to the unfiltered air inleakage test frequency following the test failure that occurred in October 2017.

The note would modify the requirements listed in TS 5.5.14.c for determining the unfiltered air inleakage past the CRE boundary into the CRE and would state:

- NOTE -

The three-year test frequency for the CRE unfiltered air inleakage test failure that occurred in October 2017 may be extended an additional three years, not to exceed October 2023.

3.0 TECHNICAL EVALUATION

In Sections 3.1 and 3.2 below, the NRC staff describes the evaluations and analyses provided by the licensee to justify the proposed changes to TSs 3.4.16, 3.7.13, 5.5.7, and 5.5.14.c.

3.1 Radiological Consequences of Design-Basis Accidents

Background

The licensee has implemented alternate source term (AST) methodology in the current licensing bases (CLB) at Beaver Valley, Units 1 and 2, through various phases. The licensee first requested a selective implementation of the AST for the fuel handling accident (FHA) by letter dated March 19, 2001 (ADAMS Accession No. ML010810433). The NRC staff approved the selective AST evaluation of the FHA radiological consequence analysis with License Amendment Nos. 241 and 121 for Units 1 and 2, respectively, dated August 30, 2001 (ADAMS Accession

No. ML012330496). Subsequently, to support the atmospheric containment conversion, the licensee requested a full implementation of the AST for the loss-of-coolant accident (LOCA) and the CREA in its LAR dated June 5, 2002 (ADAMS Accession No. ML021620298).

The NRC staff notes that the inclusion of an AST evaluation of the LOCA is a full implementation of the AST as stated in RG 1.183. The NRC staff approved the LOCA and CREA radiological consequence analyses with License Amendment Nos. 257 and 139 for Units 1 and 2, respectively, dated September 10, 2003 (ADAMS Accession No. ML032530204). In License Amendment Nos. 257 and 139, the NRC staff approved the full implementation of the AST at a thermal power level of 2,918 megawatts thermal (MWt), which included an approximate 8 percent increase to support an eventual extended power uprate (EPU) with an uncertainty of 0.6 percent for calorimetric thermal power measurements. The NRC staff had previously approved the use of 0.6 percent calorimetric thermal power measurement allowance based on reduced measurement uncertainties in License Amendment Nos. 243 and 122 for Units 1 and 2, respectively, dated September 24, 2001 (ADAMS Accession No. ML012490569).

By letter dated April 13, 2005 (ADAMS Accession No. ML051080573), the licensee requested an amendment for operation of Unit 1 with replacement steam generators (RSGs). The application included dose consequence analyses for the MSLB, SGTR, main coolant pump LRA, LACP, and the small line break (SLB) events implementing the AST and analyzed at the eventual EPU power level of 2,918 MWt (2,900 MWt plus 0.6 percent uncertainty). The NRC staff approved the MSLB, SGTR, LRA, LACP, and SLB radiological consequence analyses on February 9, 2006, with the issuance of License Amendment No. 273 for Unit 1 (ADAMS Accession No. ML060240146). For the LRA, LACP, and SLB accidents, all parameters, assumptions, accident sequences, and analysis methods used in the radiological consequence analyses are the same for both Unit 1 with the RSGs and Unit 2 with the original SGs, since bounding parameters were used to make the dose consequence analyses applicable to either unit. Therefore, the radiological consequence analyses performed at the EPU conditions for the LRA, LACP, and SLB accidents in License Amendment No. 273 for Unit 1 bound those accidents for Unit 2.

By letter dated October 4, 2004, the licensee submitted an amendment request for Units 1 and 2 to increase the rated power level from 2,689 MWt to 2,900 MWt (ADAMS Accession No. ML042920300). The application included dose consequence analyses for the Unit 2 MSLB and SGTR events and the FHA for both units. These assessments were performed using the AST and were evaluated at the proposed EPU conditions. By letter dated July 19, 2006, the NRC issued License Amendment Nos. 275 and 156 for Units 1 and 2, respectively (ADAMS Accession No. ML061720248), implementing the AST in support of the 8 percent EPU with the associated NRC staff SE (ADAMS Accession No. ML061720376).

Subsequently, the LOCA dose consequence analyses were revised due to containment sump screen modifications made to address concerns related to NUREG/CR-6771, Generic Safety Issue (GSI), GSI-191: The Impact of Debris Induced Loss of ECCS Recirculation on PWR Core Damage Frequency (ADAMS Accession No. ML022410135). These modifications were approved by the NRC in License Amendment No. 280 for Unit 1, dated October 5, 2007 (ADAMS Accession No. ML072680397), and License Amendment No. 164 for Unit 2, dated March 11, 2008 (ADAMS Accession No. ML080420549).

The current LAR includes modifications to the LOCA analysis made to address Nuclear Safety Advisory Letter, NSAL-11-5, Westinghouse LOCA Mass and Energy Release Calculation Issues, dated July 25, 2011 (ADAMS Accession No. ML13239A479). NSAL-11-5 identified six

issues that could potentially impact the plant-specific LOCA mass and energy (M&E) release calculation results that are used as input to the containment integrity response analyses.

NSAL-11-5 states that since the conservatisms in the LOCA M&E release calculation methodology offset the estimated penalties due to the combined effect of six issues, including generic errors, it was determined that a substantial safety hazard does not exist. The six issues identified in NSAL-11-5 are as follows:

1. The reactor vessel modeling did not include all the appropriate vessel metal mass available from the component drawings.

2. The reactor vessel modeling did not include all the appropriate vessel metal mass in the reactor vessel barrel/baffle region.

3. The reactor coolant pump (RCP) homologous curve input incorrectly included an absolute zero point coordinate.

4. The RCP homologous curve input incorrectly contained a sign error in a coordinate value.

5. The LOCA M&E release analysis initializes at a non-conservative (low) steam generator (SG) secondary pressure condition.

6. An error found in the EPITOME computer code (WCAP-10325-P-A methodology only) that is used to determine the M&E release rate during the long-term (i.e., post-reflood) SG depressurization phase of the LOCA transient.

Description of Revised Analyses The licensees application of proposed revisions to DBA dose consequence analyses is being used to resolve a nonconforming condition at Beaver Valley where testing resulted in CR unfiltered inleakage at a value greater than the current accident dose analysis assumption.

The licensee has proposed more restrictive changes that will reduce the allowable RCS specific activities and the secondary coolant specific activities for Unit 2 only. Existing LCO 3.4.16 restricts the specific activity of the reactor coolant to 0.35 µCi/gm DE I-131 for both units. The proposed change would reduce the LCO for Unit 2 to 0.10 µCi/gm DE I-131 while the value for Unit 1 remains unchanged at 0.35 µCi/gm DE I-131. In addition, the proposed Unit 2 pre-accident spike value would be reduced from 21 µCi/gm DE I-131 to 6 µCi/gm DE I-131 while the value for Unit 1 remains unchanged at 21 µCi/gm DE I-131.

The licensee also proposes to reduce LCO 3.7.13 governing the allowable specific activity of the secondary coolant for Unit 2 from 0.10 µCi/gm DE I-131 to 0.05 µCi/gm DE I-131 while the value for Unit 1 remains unchanged at 0.10 µCi/gm DE I-131. The NRC staff notes that the licensee used the coolant concentrations based on the current Unit1 TS LCOs for all accident evaluations except for the Unit 2 MSLB, which uses coolant concentrations based on the proposed, more restrictive LCOs for Unit 2. The Unit 2 analyses such as the Unit 2 SGTR analysis that use coolant concentrations higher than allowed by the proposed Unit 2 TSs will result in higher calculated dose consequences and are, therefore, conservative evaluations.

To support operational flexibility, the licensee has proposed to increase the allowable unfiltered air inleakage into the CRE during the three modes of operation of the CR ventilation system.

The three modes of operation include the normal mode, the isolation mode, and the emergency mode. The isolation mode is the first level of response to an abnormal condition in which the CRE is automatically isolated from the normal ventilation system. The emergency mode is entered manually when operators initiate the CR emergency filtered air intake to provide a positive pressure in the CRE, thereby reducing unfiltered inleakage. The licensees proposed changes are described below:

For the normal mode, the licensee proposed to increase the air inflow to the Beaver Valley common CR due to unfiltered intake plus inleakage from the CLB value of 500 cubic feet per minute (cfm) to a maximum of 1,250 cfm. The proposed maximum normal operation unfiltered air intake to the CR is an analytical upper bound value that is intended to include the CR ventilation intake flow rate and all unfiltered air inleakage, including a 10 cfm allowance for ingress and egress.

For the isolation mode, the licensee proposed to increase the allowable unfiltered air inleakage from the CLB maximum value of 300 cfm to 450 cfm. The updated value represents an upper bound analytical value that includes test measurement uncertainties and a 10 cfm allowance for ingress/egress.

For the emergency mode, the licensee proposed to increase the CLB maximum allowable unfiltered inleakage value of 30 cfm to 165 cfm. The updated value represents an upper bound analytical value that includes test measurement uncertainties and a 10 cfm allowance for ingress/egress.

In the LAR dated October 20, 2019, the licensee proposed to use the fission product inventory fuel gap fractions from Table 3 of DG-1199 (NRC Draft RG 1.183) when assessing the dose consequences of the Beaver Valley non-LOCA events other than reactivity-initiated accidents where only the fuel clad is postulated to be breached. The licensee stated in its LAR that this change in the CLB is intended to support flexibility in future Beaver Valley fuel management schemes and is deemed to be acceptable since Beaver Valley falls within, and intends to operate within, the maximum allowable power operating envelope for pressurized-water reactors (PWRs) shown in Figure 1 of DG-1199. The use of gap fractions from DG-1199 has been previously approved in the NRC safety evaluation report (SER) for the Diablo Canyon Nuclear Power Plant, Units 1 and 2, AST license amendments dated April 27, 2017 (ADAMS Accession No. ML17012A246).

To support the licensing and plant operation changes discussed in the LAR, the licensee revised the following accidents employing the AST as described in RG 1.183:

loss-of-coolant accident control rod ejection main steam line break accident steam generator tube rupture accident locked rotor accident loss of alternating current power fuel handling accident small line break outside containment

The NRC staff notes that the LACP and SLB events are not specifically discussed in RG 1.183.

However, the licensee incorporated the general dose assessment principles outlined in RG 1.183 to evaluate these accidents. The licensee stated that in accordance with the Beaver Valley CLB, the updated analyses continue to reflect the following:

Except for the MSLB, SGTR, LRA, and FHA analyses that are unit-specific, the remaining analyses are performed with bounding parameter values intended to encompass an event at either unit.

The Unit 2 MSLB dose analysis reflects the use of alternate repair criteria to establish a maximum allowable accident-induced SG tube leakage against which cycle-specific leakage projections are compared.

The CR dose consequence analyses associated with the Unit 1 MSLB, Unit 2 MSLB, Unit 1 SGTR, Unit 2 SGTR, and Unit 1 FHA take credit for a 30-minute CR purge after the post-accident environmental release has been terminated (or significantly reduced as in the case of the Unit 1 SGTR).

Subsequent to the CR air purges as credited above, the CR ventilation system is returned to the normal mode.

No credit is taken for the filtration capability of the supplemental leak collection and release system (SLCRS).

No credit is taken for automatic initiation of the emergency (pressurization) mode of the CREVS. Availability of CREVS when credited is based on manual initiation and assumed to be available 30 minutes after the start of the accident.

The habitability of the Beaver Valley emergency response facility (ERF), which houses the technical support center (TSC) following a LOCA, is assessed without crediting the normal or emergency ventilation systems for dose reduction. In addition, the licensee did not credit the ERF structure for shielding to reduce the direct shine from the containment or the radioactive plume.

The licensee used the Beaver Valley CLB equilibrium core inventory utilized to support this LAR, which is based on a core power level of 2,918 MWt (based on a rated thermal power of 2,900 MWt plus 0.6 percent uncertainty) and current licensed values of fuel enrichment and burnup. The methodology used to develop the core inventory was approved in the NRC staffs SER associated with License Amendment Nos. 257 and 139. The licensee did not make changes to the licensing basis core inventory used for dose consequence analyses. The design-basis core activity of isotopes significant to dose consequences as previously provided in Table 5.3.3-1 of the LAR dated June 5, 2002, for License Amendment Nos. 257 and 139 was resubmitted as Table 4.1-1 in the technical report dated August 12, 2019, included in the supplement to the LAR dated April 14, 2020.

The DBA dose consequence analyses evaluated the integrated TEDE dose at the EAB for the worst 2-hour period following the onset of the accident. The integrated TEDE doses at the outer boundary of the LPZ and the integrated dose to a Beaver Valley, Units 1 and 2, CR operator were evaluated for the duration of the accident.

The licensees dose calculation model used the committed effective dose equivalent (CEDE) from inhalation and the deep dose equivalent from external exposure from submersion due to halogens and noble gases transported to offsite locations (EAB and LPZ) and in the CR. The CEDE is calculated using the dose conversion factors (DCFs) from Environmental Protection Agency (EPA)-520/1-88-020, Federal Guidance Report No. 11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, dated September 1988, which uses the methodology provided in the International Commission on Radiological Protection, ICRP Publication 30, Limits for Intakes of Radionuclides by Workers, dated 1979-1988. As described in Sections 4.1.4 and 4.2.7 of RG 1.183, the effective dose equivalent may be used in lieu of deep dose equivalent in determining the contribution of external dose to the TEDE if the whole body is irradiated uniformly. The effective dose equivalent is calculated using the DCFs from EPA-402-R-93-081, Federal Guidance Report No. 12, External Exposure to Radionuclides in Air, Water, and Soil, dated September 1993, to determine the TEDE dose as is required for AST evaluations. To calculate the direct shine dose to the CR operator, the licensee used the DCFs from American National Standards Institute/American Nuclear Society, ANSI/ANS 6.1.1-1991, Neutron and Gamma-Ray Fluence-to-Dose Factors, dated 1991. The use of this dose methodology and DCFs is consistent with the guidance in RG 1.183 and is, therefore, acceptable to the NRC staff.

The results of the evaluations performed by the licensee, as well as the applicable dose acceptance criteria from RG 1.183, are shown in Table 1 in Section 5 of this SE. CR atmospheric dispersion factors are shown in Tables 2A through 2H in Section 5 of this SE.

Offsite atmospheric dispersion factors are shown in Table 3 and CR data and assumptions are shown in Table 4 in Section 5 of this SE. Accident-specific data and assumptions are shown in Tables 5 through 11 in Section 5 of this SE.

3.1.1 Loss-of-Coolant Accident The radiological consequence design-basis LOCA analysis is a deterministic evaluation based on the assumption of a major rupture of the primary RCS piping. The accident scenario assumes the deterministic failure of the emergency core cooling system (ECCS) to provide adequate core cooling, which results in a significant amount of core damage as specified in RG 1.183. This general scenario does not represent any specific accident sequence but is representative of a class of severe damage incidents that was evaluated in the development of the RG 1.183 source term characteristics. Such a scenario would be expected to require multiple failures of systems and equipment and lies beyond the severity of incidents evaluated for design-basis transient analyses.

The LOCA considered in this evaluation is a complete and instantaneous circumferential severance of the primary RCS piping, which would result in the maximum fuel temperature and primary containment pressure among the full range of LOCAs. Due to the postulated loss of core cooling, the fuel heats up, resulting in the release of fission products. The fission product release is assumed to occur in phases over a 2-hour period.

When using the AST for the evaluation of a design-basis LOCA for a PWR, it is assumed that the initial fission product release to the containment will last for 30 seconds and will consist of the radioactive materials dissolved or suspended in the RCS liquid. After 30 seconds, fuel damage is assumed to begin and is characterized by clad damage that releases the fission product inventory assumed to reside in the fuel gap. The fuel gap release phase is assumed to continue until 30 minutes after the initial breach of the RCS. As core damage continues, the

gap release phase ends and the early in-vessel release phase begins. The early in-vessel release phase continues for the next 1.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months . The licensee used the LOCA source term release fractions, timing characteristics, and radionuclide grouping as specified in RG 1.183 for evaluation of the AST.

In the evaluation of the design-basis LOCA radiological analysis, the licensee considered dose contributions from the following potential activity release pathways:

containment vacuum system release containment leakage directly to the atmosphere engineered safety feature (ESF) leakage reactor water storage tank (RWST) back-leakage The licensee considered the following DBA LOCA dose contributors to the CRE analysis:

contamination of the CR atmosphere by released activity direct shine dose to the CR from external and contained sources 3.1.1.1 Radiation Source Term The licensee followed all aspects of the guidance outlined in RG 1.183, Regulatory Position 3, regarding the core inventory, release fractions, and timing for the evaluation of the LOCA. The LOCA analysis assumes that iodine will be removed from the containment atmosphere by both containment sprays and natural diffusion to the containment walls. As a result of these removal mechanisms, a large fraction of the released activity will be deposited in the containment sump.

The sump water will retain soluble gaseous and soluble fission products such as iodine and cesium, but not noble gases. The guidance from RG 1.183 specifies that the iodine deposited in the sump water can be assumed to remain in solution if the containment sump pH is maintained at or above 7. In accordance with RG 1.183, since the Beaver Valley long-term sump pH is controlled to values of 7 and greater, the chemical form of the radioiodine released from the fuel is assumed to be 95 percent cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodine. Except for noble gases and elemental and organic iodine, all fission products released are assumed to be in particulate form. Therefore, the NRC staff finds the licensees assessment of the radiation source term to be consistent with RG 1.183, and therefore, acceptable.

3.1.1.2 Containment Vacuum System Release The licensee assumes that the containment vacuum system would be operating at the start of a LOCA event, providing a path for release to the environment. This line is projected to be isolated as a result of a containment isolation signal within 5 seconds of the initiating event.

Since the onset of fission product releases from the fuel is assumed to occur at 30 seconds, this pathway will be isolated prior to the onset of fuel damage. For purposes of analysis, the licensee assumes that the entire RCS inventory of volatile radionuclides is released to the containment from where it enters the environment at a rate of 2,200 cfm for 5 seconds. Since fuel damage and containment sprays will not have commenced in this period, the licensee assumes that the chemical form of the iodine released from the RCS is 97 percent elemental and 3 percent organic. The licensees assessment of the activity release via this pathway continues to demonstrate that its contribution to the site boundary and CR dose is negligible.

The NRC staff reviewed the licensees assessment and concludes that the methods and assumptions remain consistent with the CLB evaluation and, therefore, are acceptable.

3.1.1.3 Containment Leakage Release RG 1.183, Regulatory Position 3.7, states, The primary containment . should be assumed to leak at the peak pressure technical specification leak rate for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . For PWRs, the leak rate may be reduced after the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 50% of the technical specification leak rate.

Accordingly, the licensee assumed a containment leak rate of 0.1 percent per day for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , after which the containment leak rate is reduced to 0.05 percent per day for the duration of the LOCA event.

As previously stated, the licensee did not change the core inventory used in the LOCA analysis.

However, the licensee did make refinements to some of the plant design input values that are used to develop fission product removal coefficients as a result of the following:

1. The increase in the LOCA M&E release rates due to incorporation of Westinghouse NSAL-11-5.

2. The use of updated design input parameter values intended to conservatively bound Unit 1 operation, Unit 2 operation, or a combination of both, as appropriate.

The key parameters affected by the above are spray flow rates, spray initiation and termination times, spray coverage, containment pressure, containment air temperature, and steam condensation rates in containment. The NRC staff considers these changes to be refinements based on updated information as opposed to changes in methodology and the changes are, therefore, acceptable.

In accordance with the CLB, the licensee assumed that the fission products released from the fuel are assumed to mix instantaneously and homogeneously throughout the free air volume of the primary containment as it is released from the core. Consistent with the CLB, the licensee credits the impact of containment sprays as one of the primary means of fission product cleanup following a LOCA. The Beaver Valley design includes a containment quench spray system (QSS) and a containment recirculation spray system (RSS) at each of the units.

Following post-LOCA containment pressurization, the QSS is automatically initiated by the Containment Isolation Phase B signal (CIB) signal. The QSS injects cooling water from the RWST into the containment via the QSS spray headers.

Based on an assumption of a loss-of-offsite power (LOOP) coincident with the LOCA event, the QSS is assumed to be initiated at either unit by approximately 80 seconds and continues until the RWST inventory is depleted. The RSS is then initiated based on the RWST inventory low level indication. The RSS takes suction from the containment sump and provides recirculation spray inside containment via the recirculation spray headers. Consistent with the CLB, the licensee credits the RSS for up to 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months post-LOCA.

Because of the configuration of equipment and internal structures within the containment, the sprays do not cover the full containment volume. The licensee modeled the containment as being comprised of two regions: sprayed and unsprayed. Based on evaluations of spray nozzle coverage and containment arrangement in support of this LAR, the licensee reduced the CLB estimation of a spay coverage from 63 percent to the proposed value of 60 percent of the containment free volume being covered by the spray system.

The licensee stated that the containment atmosphere is mixed following a LOCA by four mechanisms: 1) momentum transfer from the fluid jet exiting the break, 2) momentum transfer from the spray droplets to the surrounding gas, 3) forced and natural convection flows within the containment atmosphere, and 4) molecular diffusion. The licensee stated that these mechanisms work together to enhance mixing within the containment to provide a homogeneous gas mixture and prevent local accumulation of fission products. Therefore, in accordance with the current CLB, the mixing rate between the effectively sprayed volume and unsprayed volume of the containment is assumed to be 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (hr), which is the default rate permitted by NUREG-0800 Section 6.5.2. The methodology used by the licensee to determine the containment mixing rate was previously approved by the NRC in License Amendment No. 280 for Unit 1 and License Amendment No. 164 for Unit 2. Therefore, the NRC staff continues to find the licensees methodology acceptable.

3.1.1.4 Emergency Core Cooling System Leakage Outside of the Containment During a DBA LOCA, some fission products released from the fuel will be carried to the containment sump via spillage from the RCS or by transport of activity from the containment atmosphere to the sump by containment sprays and natural processes, such as deposition and plateout. During the initial phases of a LOCA, safety injection (SI) and containment spray systems draw water from the RWST. The licensee assumed that at 1,200 seconds (5 minutes) after the start of the LOCA event, these systems would start to draw water from the containment sump instead of the RWST. This recirculation flow causes contaminated water to be circulated through piping and components outside of the containment where small amounts of system leakage could provide a path for the release of fission products to the environment.

The licensee assumed that all the radioiodines released from the fuel are instantaneously moved to the containment sump. Noble gases are assumed to remain in the containment atmosphere. The remaining radionuclides released from the fuel are assumed to be particulates that will not become airborne as a result of ECCS leakage. This source term assumption is conservative in that all the radioiodines released from the fuel are assumed to reside in both the containment atmosphere and containment sump. The licensee assumed that the leakage rate is two times the expected value of 5,700 cubic centimeters (cc)/hr (the bounding value for Unit 1; the Unit 2 value is 2,134 cc/hr) or 11,400 cc/hr. The licensee assumed that 10 percent of the entrained iodine activity is released to the atmosphere of the surrounding auxiliary building. The licensee assumed that this activity is exhausted without holdup, mixing, or filtration, and that the chemical form of the iodine released is 97 percent elemental and 3 percent organic. Because the licensee used data and assumptions consistent with the guidance in RG 1.183 and the CLB, the NRC staff finds the licensees assessment of the dose consequence from ECCS leakage to be acceptable.

3.1.1.5 Release from Reactor Water Storage Tank Due to Emergency Core Cooling Back-Leakage Although the RWST is isolated during recirculation, design leakage through ECCS valving provides a pathway for back-leakage of the containment sump water to the RWST. The RWST is located in the plant yard and is vented to the atmosphere. Since this release path represents a bypass of the containment, dose consequences must be considered. The concentration of radionuclides in the containment sump water is as modeled above for ECCS leakage. The licensee assumed that containment sump water leaks into the RWST at a rate of 2 gallons per minute (gpm), which is twice the surveillance limit of 1 gpm. The licensee assumed that the

leakage into the RWST begins at about 30 minutes (1,768 seconds) post-LOCA. The licensee assumed that radioiodines are projected to be released via the RWST vent starting at about 50 minutes (3,039 seconds) post-LOCA and continues until the end of the accident analysis period of 30 days.

The licensee assumed that a portion of the iodine dissolved in the back-leakage will be retained within the RWST. The time-dependent iodine release rates used by the licensee are illustrated in Figure 7.2-3 and Table 7.2-3 of the technical report dated August 12, 2019, and were included in the April 14, 2020, supplement to this LAR. Values range from about 0.01 per day at 3,039 seconds post-accident to a minimum value of about 2.0 x 10-7 per day at 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months . The licensee assumed that this activity is exhausted without filtration and that the chemical form of the iodine released is 97 percent elemental and 3 percent organic. The methodology used by the licensee to determine the RWST release fractions was previously approved by the NRC in License Amendment No. 280 for Unit 1 and for License Amendment No. 164 for Unit 2.

3.1.1.6 Control Room Occupancy Dose The licensees assumptions used for the evaluation of CR habitability are discussed in Section 3.1.9 of this SE and presented in Table 4 in Section 5 of this SE. Provided below are the critical LOCA-specific assumptions associated with CR response and activity transport.

In accordance with the Beaver Valley CLB, due to the rapid pressure transient expected following a LOCA, the Containment Isolation Phase B (CIB) signal, which initiates the CR isolation and emergency ventilation following a LOCA is assumed to actuate at t = 0 hours0 days 0 hours 0 weeks 0 months .

Considering a LOOP, the maximum estimated delay in attaining CR isolation after receipt of a CIB signal to switch CR ventilation from normal mode to emergency mode is 77 seconds, which accounts for delays due to emergency diesel generator (EDG) start and load sequencing (including time for damper movement/realignment).

No credit is taken for automatic initiation of the Unit 2 CREVS; rather, it is assumed the CREVS will be initiated by manual operator action providing a pressurized CR within 30 minutes of accident initiation.

The licensee evaluated the following LOCA sources that could potentially impact the CR operator dose due to direct shine:

direct shine from the containment structure direct shine from the airborne source located in the cable spreading room below the Unit 2 portion of the combined CR through floor penetrations direct shine from the airborne source located in the cable tray mezzanine area below the Unit 1 portion of the combined CR through floor penetrations direct shine through the CR walls and floor from the CREVS intake filters located in the adjacent room (Unit 2 filters) and below (Unit 1 filters), respectively

direct shine from the sump fluid that is postulated to collect in the RWST direct shine from the contaminated cloud outside the CR pressure boundary resulting from containment leakage, ESF system leakage, and RWST back-leakage The NRC staff reviewed the licensees analysis presented in Calculation No. 10080-UR(B)-487, Revision 3, Site Boundary, Control Room and Emergency Response Facility Doses following a Loss-of-Coolant Accident based on Core Uprate, an Atmospheric Containment and Alternative Source Terms, provided in the supplement to the LAR dated June 9, 2020. In Revision 3 of Calculation No. 10080-UR(B)-487 for the postulated LOCA, the licensee considered the following: a) an increase in allowable unfiltered inleakage into the CRE from 30 cfm to 165 cfm, b) the presence of wall penetrations between the Unit 2 CR filter cubicle and the CRE, c) an investigation/demonstration that the current model that does not address the presence of particulate air filters (intended for dust removal) in the control room ventilation system recirculation air conditioning system remains bounding, d) a bounding approach with respect to the ERF ventilation filters when addressing the direct shine dose, e) worst-case 30-day integrated dose estimates in (i) the ERF and (ii) the TSC, f) include a dose contribution to ERF/TSC personnel for the 30-day duration of the accident, and g) review/update (as needed) of all design input parameter values/references to reflect current plant design.

The licensees contractor (WECTEC Global Project Services, Inc.) performed the analysis using the PERC2 proprietary computer code for calculating the airborne dose to the CR operator and a member of the public located at the EAB and LPZ. In accordance with the CLB, the PERC code utilizes an exact solution analytical computational process that addresses radionuclide progeny, time-dependent releases, and transport rates between regions, and deposition of radionuclide concentrations in sumps, walls, and filters. The WECTEC PERC2 code utilizes inputs that describe the radionuclide release rates from the core (for use in AST analyses) and the path and rate of exchange between regions, including filter efficiencies, deposition rates, atmospheric dispersion, breathing rates, dose conversion, and occupancy factors. Also, in accordance with the CLB, the licensees spray model for the LOCA analysis is performed using the WECTEC SWNAUA proprietary code. The WECTEC SWNAUA code calculates the aerosol particle removal rate from containment atmosphere by the quench and recirculation spray systems.

The licensee stated that the methodology used to assess post-LOCA fission product transport and associated dose consequences remains unchanged from that discussed in the licensees amendment request dated June 5, 2002 (ADAMS Accession No. ML021620298) (licensees LAR Nos. 300 and 172), as amended by the licensees amendment request dated February 9, 2007 (ADAMS Accession No. ML070440341) (licensees LAR Nos. 334 and 205). The licensees methodology was previously approved by the NRC staff in SERs associated with License Amendment Nos. 257 and 139 implementing the AST, License Amendment No. 280 for Unit 1 and License Amendment No. 164 for Unit 2. Also, the licensee stated that the spray model utilized for the LOCA for the analysis remains unchanged from that discussed in the licensees LAR Nos. 300 and 172, and was approved in License Amendment Nos. 257 and 139.

The methodology was further approved by the NRC staff for the licensees LAR Nos. 334 and 205 and in the NRC staffs SERs for License Amendment Nos. 280 and 164, respectively. The licensee concluded from its analysis that the Beaver Valley site boundary and CR doses due to radioactive material released following a LOCA will remain within the regulatory limits set by 10 CFR 50.67 and discussed in RG 1.183, and that the ERF doses also remain within 5 rem TEDE. Therefore, the NRC staff continues to find the licensees methodology acceptable.

3.1.1.7 Loss-of-Coolant Accident Conclusion Based on the above, the NRC staff concludes that the licensees analysis was performed using acceptable models, assumptions, and inputs consistent with applicable regulatory guidance identified in Section 2.0 of this SE. The assumptions and inputs that the NRC staff found acceptable are presented in Table 5 in Section 5 of this SE, and the licensees calculated dose results are given in Table 1 in Section 5 of this SE. The LOCA atmospheric dispersion factors previously accepted by the NRC staff are shown in Table 2A in Section 5 of this SE. The staff reviewed the licensees evaluation of the radiological consequences resulting from the postulated LOCA and concludes that the radiological consequences at the EAB, LPZ, and CR continue to meet the acceptance criteria provided in 10 CFR 50.67 and accident-specific acceptance criteria specified in SRP Section 15.0.1. Therefore, the licensees LOCA evaluation is acceptable from a dose consequence perspective.

3.1.2 Control Rod Ejection Accident This event consists of an uncontrolled withdrawal of a control rod from the reactor core. The CREA results in reactivity insertion that leads to a core power level increase, resulting in fuel damage and a subsequent reactor trip. Following the reactor trip and based on an assumption of a LOOP coinciding with the reactor trip, the condenser is assumed to be unavailable, and reactor cooldown is achieved using steam releases from the SG main steam safety valves (MSSVs) and atmospheric dump valves (ADVs). Since the residual heat removal (RHR) pumps are located inside containment and are not qualified for Containment Isolation Phase B (CIB) conditions, which are expected to occur following a CREA, no credit is taken for initiation of shutdown cooling, and environmental releases via the MSSVs/ADVs are assumed to occur for 30 days.

The CLB assumptions for the CREA were previously approved by the NRC for License Amendment Nos. 257 and 139. The licensee used the guidance in Appendix H of RG 1.183 to develop the dose consequences following a postulated CREA. Table 6 in Section 5 of this SE lists some of the key data and assumptions used by the licensee to develop the radiological consequences following the CREA. The licensee used bounding values to encompass the event occurring at either unit.

Under RG 1.183, which the licensee followed, the CREA evaluation addresses two independent release paths to the environment. The first release path evaluated is the containment leakage pathway, which assumes that the released fission products breach the RCS and are released into the containment. The second release pathway assumes that the RCS remains intact and that secondary side releases occur from the assumed primary-to-secondary leakage in the SGs.

The licensee notes, and the NRC staff agrees, that the actual doses resulting from a postulated CREA would be a composite of the releases from the containment leakage pathway and the secondary system release pathway. Therefore, if the evaluation indicates the acceptance criteria is met for each of the independent scenarios, then the dose consequence of a scenario that is a combination of the two will be encompassed by the more restrictive of the two analyzed scenarios.

In accordance with the CLB, the licensee assumed that the CREA results in less than 10 percent fuel cladding failure, resulting in the release of the associated fuel gap activity and less than 0.25 percent fuel centerline melt. The CREA evaluation includes the use of a radial peaking factor to account for differences in power level across the core to determine the isotope

inventory of the damaged rods. The licensee modified the radial peaking factor from the CLB value of 1.75 to a value of 1.7, which continues to bound the current design value of 1.62 per the Core Operating Limits Reports (COLRs) in the Beaver Valley, Unit 1, Licensing Requirements Manual (LRM), Revision 102, and Beaver Valley, Unit 2, LRM, Revision 94.

The licensee followed the guidance in RG 1.183 by assuming that 10 percent of the core inventory of the noble gases and iodines resides in the fuel gap and is released based on the fraction of fuel cladding failure. For the containment leakage pathway, following RG 1.183 guidance, the licensee assumed that the activity attributed to fuel melting is based on the fraction of the fuel that reaches or exceeds the initiation temperature for fuel melting, and the assumption that 100 percent of the noble gases and 25 percent of the iodines contained in that fraction are available for release from containment. For the secondary system release pathway, following RG 1.183 guidance, the licensee assumed 100 percent of the noble gases and 50 percent of the iodines in the fuel melt fraction are released to the reactor coolant.

Since the licensee did not credit actuation of the CIB signal, the CR emergency ventilation system is assumed to be manually initiated 30 minutes after the CREA. The remaining CR parameters incorporated in the licensees analysis are shown in Table 4 in Section 5 of this SE.

3.1.2.1 Control Rod Ejection Accident Conclusion The NRC staff concludes 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 Table 6 in Section 5 of this SE, and the licensees calculated dose results are given in Table 1 in Section 5 of this SE. The CREA atmospheric dispersion factors previously accepted by the NRC staff are shown in Table 2B in Section 5 of this SE. The NRC staff reviewed the licensees evaluation of the radiological consequences resulting from the postulated CREA and concludes that the radiological consequences at the EAB, LPZ, and CR continue to meet the acceptance criteria provided in 10 CFR 50.67 and the accident-specific acceptance criteria specified in SRP Section 15.0.1. Therefore, the licensees CREA evaluation is acceptable with respect to its dose consequence analysis.

3.1.3 Main Steam Line Break The MSLB event consists of a double-ended break of one main steam line. The analysis focuses on an MSLB outside the containment, since an MSLB inside containment results in a lesser dose to a CR operator or to the public at offsite locations due to holdup of activity in the containment. Following an MSLB outside containment, the faulted SG rapidly depressurizes and releases its initial secondary side liquid contents to the environment via the break. Based on an assumption of a simultaneous LOOP, the condenser is assumed to be unavailable, and environmental steam releases via the MSSVs/ADVs of the intact SGs are used to cool down the reactor until initiation of shutdown cooling via the RHR system.

The NRC staff previously evaluated the radiological consequences for the MSLB and SGTR accidents for Unit 1 in License Amendment No. 273 in support of Unit 1 operation with the RSGs at a reactor core power level of 2,918 MWt and implementing the AST. The NRC staff previously evaluated the radiological consequences for the MSLB and SGTR accidents for Unit 2 in License Amendment Nos. 275 and 156 in support of the Beaver Valley EPU.

The licensees CLB analysis indicates that no fuel melting or fuel cladding failure is postulated for the Beaver Valley MSLB event. Therefore, in accordance with RG 1.183, the licensee assumed that the activity released is based on the maximum coolant activity allowed by the

Beaver Valley TSs. Following the guidance in RG 1.183, the licensee considered two spiking scenarios - a pre-accident iodine spike, which reflects the maximum allowable TS iodine spike activity level, and an accident-initiated iodine spike, which results in an increase in the iodine appearance rate from the fuel to the RCS by a factor of 500.

In accordance with the CLB, the licensees MSLB assessment for Unit 2 supports the implementation of alternate repair criteria as defined in NRC Generic Letter, GL 95-05, Voltage Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking, dated August 3, 1995. In accordance with GL 95-05, an accident-induced primary-to-secondary leakage is postulated to occur (via preexisting tube defects) as a result of the rapid depressurization of the secondary side due to the postulated MSLB and the consequent high differential pressure across the faulted SG. The maximum allowable accident-induced SG tube leakage rate is the maximum primary-to-secondary leakage that could occur in the faulted SG with the offsite or CR operator doses remaining within applicable limits. The Unit 2 MSLB analysis assigns this tube leakage to the faulted SG.

Consequently, the primary-to-secondary leakage in the faulted SG reflects 150 gallons per day (gpd) at standard temperature and pressure (STP), plus the maximum allowable accident-induced tube leakage that results in dose consequences that are within the most limiting of the regulatory limits associated with the EAB, LPZ, and CR.

The current LAR proposes to increase the allowable accident-induced primary-to-secondary SG tube leakage limit from the CLB value of 2.1 gpm to 8.1 gpm. In addition, the licensee is proposing to lower the Unit 2 reactor and secondary coolant TS activity concentrations to 0.10 Ci/gm DE I-131 and 0.05 Ci/gm DE I-131, respectively.

The licensee credited manual initiation of the CREVS at 30 minutes following the event, resulting in CR pressurization with the accompanying reduction in unfiltered inleakage. In addition, the licensee credited a CR purge at a rate of 16,200 cfm for a period of 30 minutes beginning at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the MSLB event.

3.1.3.1 Main Steam Line Break Conclusion The NRC staff concludes 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 Table 7A for Unit 1 and Table 7B for Unit 2 in Section 5 of this SE, and the licensees calculated dose results are given in Table 1 in Section 5 this SE. The MSLB atmospheric dispersion factors previously accepted by the NRC staff are shown in Table 2C in Section 5 of this SE. The staff reviewed the licensees evaluation of the radiological consequences resulting from the postulated MSLB and concludes that the radiological consequences at the EAB, LPZ, and CR continue to meet the acceptance criteria provided in 10 CFR 50.67 and the accident-specific acceptance criteria specified in SRP Section 15.0.1. Therefore, the licensees MSLB evaluation is acceptable with respect to its dose consequence analysis.

3.1.4 Steam Generator Tube Rupture This event assumes the instantaneous rupture of an SG tube with a resultant release of reactor coolant into the lower pressure secondary system. Based on an assumption of a LOOP occurring simultaneously with the reactor trip, the condenser is assumed to be unavailable, and environmental steam releases via the MSSVs/ADVs of the intact SGs are used to cool down the reactor until initiation of shutdown cooling via the RHR system. A portion of the reactor coolant

break flow into the ruptured SG flashes and is released to the condenser for a short duration prior to the reactor trip, and thereafter, directly to the environment via the MSSVs/ADVs. The remaining break flow mixes with the secondary side liquid and is released to the environment via steam releases through MSSVs/ADVs. The activity in the RCS also leaks into the intact SGs via SG tube leakage and is released to the environment from the MSSVs/ADVs.

The NRC staff previously evaluated the radiological consequences for the MSLB and SGTR accidents in License Amendment No. 273 for Unit 1 in support of Unit 1 operation with the RSGs at the anticipated EPU reactor core power level of 2,918 MWt and implementing the AST.

The NRC staff previously evaluated the radiological consequences for the MSLB and SGTR accidents for Unit 2 in License Amendment Nos. 275 and 156 in support of the Beaver Valley EPU.

Appendix F of RG 1.183 identifies acceptable radiological analysis assumptions for an SGTR.

The licensees analysis indicates that no fuel melting or fuel cladding failure is postulated for the SGTR event. Therefore, in accordance with RG 1.183, the licensee assumed that the activity released is based on the maximum coolant activity allowed by the Unit 1 TSs. The staff notes that the licensee used the higher Unit 1 TS values of allowable coolant activity for both the Unit 1 and 2 SGTR assessments. Therefore, the Unit 2 SGTR analysis has added conservatism since the lower coolant values, as governed by the proposed Unit 2 TSs, are not credited in the analysis. Following the guidance in RG 1.183, the licensee considered two spiking scenarios - a pre-accident iodine spike, which reflects the maximum allowable Unit 1 TS iodine spike activity level, and an accident-initiated iodine spike, which results in an increase in the iodine appearance rate from the fuel to the RCS by a factor of 335.

Primary-to-secondary leakage is assumed to be 150 gpd into the bulk water of the ruptured SG and 300 gpd total into the bulk water of the two intact SGs, as permitted by the Unit 2 TSs. The iodine activity from the flashed portion of the break flow through the ruptured SG is assumed to be directly released to the environment with no iodine partitioning. The radionuclides in the intact SG bulk water are assumed to become vapor at a rate that is a function of the steaming rate for the SGs and the partition coefficient. The licensee assumed that the radionuclide concentration in the SG is partitioned such that 1.0 percent of the radionuclides in the unaffected SG bulk water enter the vapor space and are released to the environment. The steam release from the unaffected SGs continues for approximately 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months until the RHR shutdown cooling system can be used to complete the cooldown.

The licensee claimed no credit for fission product removal by the CREVS following an SGTR event and assumed the CR is maintained in normal ventilation mode. Following termination of the environmental release at 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , the CR is purged at a rate of 16,200 cfm for a period of 30 minutes.

The NRC staff reviewed the licensees analysis presented in Calculation No. 8700-UR(B)-219, Revision 3, Site Boundary and Control Room Doses following a Steam Generator Tube Rupture (SGTR) based on Core Uprate and Alternative Source Term (Unit 1), and Calculation No. 10080-UR(B)-496, Revision 3, Site Boundary and Control Room Doses following a Steam Generator Tube Rupture based on Core Uprate and Alternative Source Term Methodology (Unit 2), provided in the supplement to the LAR dated June 9, 2020. In Revision 3 of Calculation No. 8700-UR(B)-219 and Calculation No. 10080-UR(B)-496 for the postulated SGTR, the licensee considered: a) an increase in allowable unfiltered inleakage into the CRE (inclusive of that associated with ingress /egress) from 30 cfm to 165 cfm and b) review/update (as needed) of all design input parameter values/references to reflect current plant design.

As described in Section 3.1.1.6 of this SE, the PERC2 code is used to calculate the CR and site boundary doses due to airborne radioactivity releases following SGTRs at Units 1 and 2. The licensee stated that the CLB thermal hydraulic analysis model used to determine the dose consequences at the site boundary and CR for the Unit 1 SGTR is a simplified transient model (a conservative hand calculation method), which was the common industry standard prior to 1980. In support of the EPU, the licensee performed an assessment to demonstrate that the dose estimates using the CLB thermal hydraulic analysis model bounded that developed based on a more realistic SGTR transient analysis, which considered realistic operator action times, single failure, margin for SG overfill, and a 10-minute stuck-open ADV. In accordance with the CLB, the analysis addresses reduced capacity of the ADVs in that environmental releases are assumed to continue between t = 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to t = 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The licensee stated that with the exception of the Unit 2 MSLB, the Unit 1 equilibrium TS iodine activity concentration of 0.35 Ci/gm DE I-131 for the reactor coolant and 0.10 Ci/gm DE I-131 for the secondary coolant will be conservatively used in all Unit 2 dose consequence analyses that are based on release of coolant at TS concentrations with or without iodine spikes. To assess the dose consequences at the site boundary and CR for the Unit 2 SGTR, the licensee utilized the same CLB thermal hydraulic analysis model as for the Unit 1 SGTR, which takes into consideration realistic operator action times, single failure, margin for SG overfill, and a 10-minute stuck-open ADV. The licensee notes that the steam release from the faulted SG includes a short period release between 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months when the faulted SG is manually depressurized in preparation for RHR operation.

The licensees methodology used to assess the SGTRs at Units 1 and 2 and associated dose consequences are discussed in LAR Nos. Nos. 302 and 173 for the EPU, and was approved in the NRC staffs SER for License Amendment Nos. 275 and 156. The licensee concluded from its analysis that the Beaver Valley site boundary and CR doses due to airborne radioactive material released following SGTRs at Units 1 and 2 will remain within the regulatory limits set by 10 CFR 50.67 and discussed in RG 1.183.

3.1.4.1 Steam Generator Tube Rupture Conclusion Based on the above, the NRC staff concludes that the licensees analysis was performed using acceptable models, assumptions, and inputs performed consistent with applicable regulatory guidance identified in Section 2.0 of this SE. The assumptions found acceptable to the NRC staff are presented in Table 8A for Unit 1 and Table 8B for Unit 2 in Section 5 of this SE, and the licensees calculated dose results are given in Table 1 in Section 5 of this SE. The SGTR atmospheric dispersion factors previously accepted by the NRC staff are shown in Table 2D in Section 5 of this SE. The staff reviewed the licensees evaluation of the radiological consequences resulting from the postulated SGTR and concludes that the radiological consequences at the EAB, LPZ, and CR continue to meet the acceptance criteria provided in 10 CFR 50.67 and the accident-specific acceptance criteria specified in SRP Section 15.0.1.

Therefore, the licensees SGTR evaluation is acceptable with respect to its dose consequence analysis.

3.1.5 Main Coolant Pump (Reactor Coolant Pump) Locked Rotor Accident The accident considered is the instantaneous seizure of an RCP rotor, which causes a rapid reduction in the flow through the affected RCS loop. For this accident scenario, a reactor trip occurs, SI actuates, and a LOOP occurs concurrently with the reactor trip. The flow imbalance creates localized temperature and pressure changes in the core. The radiological

consequences are due to leakage of the radioactive reactor primary coolant to the SGs and from there to the environment. Because the LOOP renders the main condenser unavailable, the plant is cooled down by release of steam to the environment through ADVs and MSSVs.

The releases to the environment are assumed to continue for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , at which time shutdown cooling is initiated via operation of the RHR system. Appendix G of RG 1.183 identifies acceptable radiological analysis assumptions for an LRA.

The licensee assumed that due to the LRA rendering the affected RCP inoperable, the resulting loss of primary coolant circulation may result in as much as 20 percent of the core fuel rods experiencing a departure from nucleate boiling. This will cause fuel cladding damage and release of the gap activity in the damaged fuel into the RCS.

In its LAR dated October 20, 2019, the licensee proposed to use the fission product inventory fuel gap fractions from Table 3 of NRC DG-1199 when assessing the dose consequences of Beaver Valley non-LOCA events other than reactivity-initiated accidents where only the fuel clad is postulated to be breached. The licensee stated in its LAR that this change in licensing basis is intended to support flexibility in future Beaver Valley fuel management schemes and is deemed to be acceptable since Beaver Valley falls within, and intends to operate within, the maximum allowable power operating envelope for PWRs shown in Figure 1 of DG-1199. The use of gap fractions from DG-1199 has been previously approved in the NRC staffs SER for the Diablo Canyon Nuclear Power Plant, Units 1 and 2, AST license amendments dated April 17, 2017.

In addition, the licensee proposed to change the radial peaking factor from the CLB value of 1.75 to 1.7. In the technical report dated August 12, 2019, submitted with the April 14, 2020, supplement to the LAR, the licensee states that the reduction in the peaking factor margin to 1.7 continues to be bounded by the radial peaking factor, which is currently designed to 1.62 per the COLRs in Beaver Valley, Unit 1, LRM, Revision 102, and Beaver Valley, Unit 2, LRM, Revision 94.

The radionuclides released from the fuel are assumed to be instantaneously and homogeneously mixed in the RCS and transported to the secondary side via primary-to-secondary leakage of 450 gpd for all three SGs for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The licensee assumed that this leakage mixes with the bulk water of the SGs secondary side, and that the radionuclides in the bulk water become vapor at a rate that is a function of the steaming rate for the SGs and the partition coefficient. The licensee computed the CR doses for this event assuming that the CR remained in the normal ventilation mode for the duration of the accident.

As described in Sections 3.1.1.6 and 3.1.4 of this SE, the PERC2 code is used for calculating the airborne dose to the CR operator and to a member of the public located at the EAB and LPZ. The NRC staff reviewed the licensees analysis presented in Calculation No. 10080-UR(B)-493, Revision 1, Site Boundary and Control Room Doses based on Core Uprate and Alternative Source Term Methodology following a) a Locked Rotor Accident b) a Loss of AC Power Accident, in the supplement to the LAR dated June 9, 2020. In Revision 1 of Calculation No. 10080-UR(B)-493 for the postulated LRA, the licensee considered: a) an increase in allowable unfiltered inleakage into the CRE (inclusive of that associated with ingress/egress) from 30 cfm to 165 cfm, b) use of fuel gap fractions from Table 3 of DG-1199 for all non-LOCA events that experience fuel damage with the exception of the CREA, and c) review/update (as needed) of all design input parameter values/references to reflect current plant design. The licensee concluded from its analysis that the Beaver Valley site boundary and

CR doses due to airborne radioactive material released following an LRA will remain within the regulatory limits set by 10 CFR 50.67.

3.1.5.1 Locked Rotor Accident Conclusion Based on the above, the NRC staff concludes that the licensees analysis was performed using acceptable models, assumptions, and inputs consistent with applicable regulatory guidance identified in Section 2.0 of this SE. The assumptions and inputs found acceptable to the NRC staff are presented in Table 9 in Section 5 of this SE, and the licensees calculated dose results are given in Table 1 in Section 5 of this SE. The LRA atmospheric dispersion factors previously accepted by the NRC staff are shown in Table 2E in Section 5 of this SE. The NRC staff reviewed the licensees evaluation of the radiological consequences resulting from the postulated LRA and concludes that the radiological consequences at the EAB, LPZ, and CR continue to meet the acceptance criteria provided in 10 CFR 50.67, the accident-specific acceptance criteria specified in SRP Section 15.0.1. Therefore, the licensees LRA evaluation is acceptable with respect to its dose consequence analysis.

3.1.6 Loss of (Non-Emergency) Alternating Current Power The LACP to the station auxiliaries accident is the result of a complete loss of either the external (offsite) grid or the onsite alternating current (AC) distribution system. All RCPs are tripped simultaneously by the initiating event, resulting in a flow coastdown, as well as a decrease in heat removal by the secondary system. The LACP results in the main condenser becoming unavailable, a reactor trip, and reactor cooldown being achieved using steam releases via the MSSVs/ADVs until initiation of shutdown cooling.

The licensee stated, and the NRC staff agrees, that the plants response to a postulated LACP event is similar to the response to an LRA with the exception that the LRA event results in fuel cladding damage and associated release of gap activity, whereas the LACP event involves no fuel damage. Therefore, since the available plant systems to deal with the events are similar but the source term for the LRA is significantly larger than for an LACP, the radiological consequences resulting from the LRA event will bound the consequences from the LACP event.

3.1.6.1 Loss of Alternating Current Power Conclusion The NRC staff concludes that the licensee considered analysis assumptions and inputs consistent with applicable regulatory guidance identified in Section 2.0 of this SE. The assumptions and inputs found acceptable to the NRC staff are presented in Table 10 in Section 5 of this SE. The LACP atmospheric dispersion factors previously accepted by the NRC staff are shown in Table 2F in Section 5 of this SE. Consistent with the CLB, the licensee did not compute accident-specific doses for the LACP since the dose consequences from the LRA bound the dose consequences from an LACP. The NRC staff concludes that the radiological consequences at the EAB, LPZ, and CR continue to meet the acceptance criteria provided in 10 CFR 50.67 and the most restrictive accident-specific acceptance criteria specified in SRP Section 15.0.1. Therefore, the licensees LACP evaluation is acceptable from a dose consequence perspective.

3.1.7 Fuel Handling Accident The licensee previously analyzed the FHA at a reactor power level of 2,705 MWt implementing the AST and submitted it for the NRC approval by letter dated March 19, 2001. The NRC staff

approved the FHA radiological consequence analysis with License Amendment Nos. 241 and 121. Subsequently, the licensee reanalyzed the radiological consequence of an FHA in support of an EPU. The updated FHA analysis assumed a core power of 2,918 MWt, which reflects the EPU power level of 2,900 MWt with an additional 0.6 percent to account for measurement uncertainty. By letter dated July 19, 2006, the NRC staff approved the reanalyzed FHA radiological consequence analysis in License Amendment Nos. 275 and 156 for Units 1 and 2, respectively. In both applications, the NRC staff concluded that the licensee used analysis methods and assumptions consistent with the guidance described in RG 1.183.

The FHA assumes the dropping of a single spent fuel assembly in the fuel pool area located in the fuel building or in the reactor cavity located inside containment during refueling. The Unit 1 and Unit 2 TS LCO 3.9.3, Refueling Operations - Decay Time, requires the reactor to be subcritical for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to moving irradiated fuel. This requirement prohibits initiation of fuel handling activities in the fuel pool or in the containment until 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after reactor shutdown. The FHA assumes 137 of the 264 fuel rods in a fuel assembly are damaged, and the radionuclide inventory in the fuel rod gap of the damaged fuel rods is assumed to be instantaneously released to either the water in the reactor cavity or the spent fuel pool water.

These assumptions are aligned with the current Beaver Valley design and licensing bases, which are documented in the Units 1 and 2 Updated Final Safety Analysis Reports (UFSARs) and were previously accepted by the NRC in License Amendment Nos. 241 and 121 and License Amendment Nos. 275 and 156.

In the LAR dated October 20, 2019, the licensee proposed to use the fission product inventory fuel gap fractions from Table 3 of DG-1199 when assessing the dose consequences of Beaver Valley non-LOCA events other than reactivity-initiated accidents where only the fuel clad is postulated to be breached. The licensee stated in its LAR that this change in licensing basis is intended to support flexibility in future Beaver Valley fuel management schemes and is deemed to be acceptable since Beaver Valley falls within, and intends to operate within, the maximum allowable power operating envelope for PWRs shown in Figure 1 of DG-1199. The use of gap fractions from DG-1199 has been previously approved in the NRC staffs SER for the Diablo Canyon Nuclear Power Plant, Units 1 and 2, AST license amendments dated April 17, 2017.

In addition, the licensee proposed to change the radial peaking factor from the CLB value of 1.75 to 1.7. In the technical report dated August 12, 2019, submitted with the April 14, 2020, supplement to the LAR, the licensee states that the reduction in the peaking factor margin to 1.7 continues to be bounded by the radial peaking factor, which is currently designed to 1.62 per the COLRs in Beaver Valley, Unit 1, LRM, Revision 102, and Beaver Valley, Unit 2, LRM, Revision 94.

The licensee assumed all fission products released from the damaged fuel rods are released to the fuel pool or reactor cavity. The licensee further assumed no decontamination for noble gases, an effective decontamination factor of 200 for radioiodines, and retention of all aerosol and particulate radionuclides within the spent fuel pool or reactor cavity water. The licensee then assumed that all noble gases and iodine from the spent fuel pool or reactor cavity are released from either an open containment or an open fuel building to the environment in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months without any credit for filtration, holdup, or dilution. Appendix B of RG 1.183 identifies acceptable radiological analysis assumptions for an FHA.

The licensee did not credit fission product removal by the CREVS following an FHA and assumed the CR is maintained in normal ventilation mode. For the Unit 1 FHA analysis, the

licensee credited a CR purge at a rate of 16,200 cfm for a period of 30 minutes following the termination of environmental releases at 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months post-FHA.

3.1.7.1 Fuel Handing Accident Conclusion The NRC staff concludes 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 Table 11 in Section 5 of this SE and the licensees calculated dose results are given in Table 1 in Section 5 of this SE. The FHA atmospheric dispersion factors previously accepted by the NRC staff are shown in Table 2G in Section 5 of this SE. The NRC staff reviewed the licensees evaluation of the radiological consequences resulting from the postulated FHA and concludes that the radiological consequences at the EAB, LPZ, and CR continue to meet the acceptance criteria provided in 10 CFR 50.67 and the accident-specific acceptance criteria specified in SRP Section 15.0.1. Therefore, the licensees FHA evaluation is acceptable with respect to its dose consequence analysis.

3.1.8 Small Line Break Outside Containment The CLB SLB event postulates the break of a 2-inch RCS letdown line in the auxiliary building outside of the containment, resulting in a maximum break flow of 16.79-pound mass per second (lbm/s). The letdown line is the largest piping that carries RCS fluid outside containment. A rupture of the letdown line provides a release path for the primary coolant to the environment through the auxiliary building ventilation vent. The break flow rate is based on subcooled critical flow, taking into consideration choked flow via two letdown orifices. The iodine activity in the break flow is assumed to become airborne in proportion to the flash fraction, whereas the noble gases are assumed to be airborne and discharged to the environment without decontamination or holdup. The flash fraction of 37 percent assumed in the analysis corresponds to an assumption that the fluid is being released at reactor coolant temperature conditions. The licensee did not credit any cooling of the fluid as a result of the regenerative heat exchanger, which is located inside the containment. The licensee used bounding parameter values to encompass an event at either unit.

In accordance with the CLB, the licensee credits manual operator action from the CR to isolate the break within 15 minutes of accident initiation. The licensee notes that if the break was large enough to cause an SI signal, the letdown line would be automatically isolated; however, the SLB analysis assumes the break would not trigger an SI signal. The licensees analysis indicates that no fuel failure results from the letdown line break, which is consistent with the CLB. The radioactivity in the RCS is assumed to be at the bounding equilibrium iodine TS limit of 0.35 Ci/gm DE I-131 for Unit 1. The consideration of the equilibrium iodine TS limit of 0.35 Ci/gm DE I-131 is consistent with the review procedure provided in SRP Section 15.6.2, Radiological Consequences of the Failure of Small Lines Carrying Primary Coolant Outside Containment. Neither RG 1.183 nor SRP 15.0.1 address the SLB event as a DBA. The accident was assumed to cause the iodine concentration to spike by a factor of 500 times the equilibrium iodine appearance rate for a period of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . A total of 15,110 lbm of RCS fluid was assumed released through the break based on a break mass flow rate of 16.79 lbm/s for 15 minutes. The break flow rate is based on subcooled critical flow, taking into consideration choked flow via two letdown orifices. The licensee assumed 37 percent of the break flow would flash based on a constant enthalpy process. These parameters are consistent with the CLB.

The iodine activity in the break flow is assumed to be airborne in proportion to the flash fraction, whereas the noble gases are assumed to be airborne and released to the environment without

decontamination or holdup. The chemical form of the iodine released to the environment is assumed to be 97 percent elemental iodine and 3 percent organic iodide.

The SLB event is not expected to result in an SI signal. Therefore, the licensee assumed no isolation of the CR. The analysis for the SLB event assumes that the CR remains in its normal operation mode with a maximum unfiltered air intake of 1,250 cfm for the accident duration.

RG 1.183 does not address an SLB outside containment. The licensees analysis used assumptions and inputs that follow the guidance provided for similar DBAs in RG 1.183 and the SRP. Since there are no specific dose acceptance criteria given in SRP Section 15.0.1 for the letdown line break, the licensee used the most limiting dose acceptance criteria for any DBA listed in RG 1.183 (2.5 rem TEDE at the EAB and the outer boundary of the LPZ and 5 rem TEDE in the CR). The NRC staff found 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 Table 12 in Section 5 of this SE, and the licensees calculated dose results are given in Table 1 in Section 5 of this SE. The SLB atmospheric dispersion factors previously accepted by the NRC staff are shown in Table 2H in Section 5 of this SE. The NRC staff reviewed the licensees evaluation of the radiological consequences resulting from the postulated SLB and concludes that the radiological consequences at the EAB, LPZ, and CR continue to meet the acceptance criteria provided in 10 CFR 50.67 and the most restrictive accident-specific acceptance criteria specified in SRP Section 15.0.1. Therefore, the licensees SLB evaluation is acceptable from a dose consequence perspective.

3.1.9 Control Room Envelope Unfiltered Air Inleakage The licensee states that Beaver Valley is served by a single CR with two ventilation intakes, each assigned to Units 1 and 2, which are used during the normal ventilation and the emergency ventilation modes of operation.

The licensee states that to provide operational margin, the dose consequence analyses provided in support of this LAR assume that during normal plant operation, the joint Units 1 and 2 unfiltered intake plus inleakage would be a maximum of 1,250 cfm (from 500 cfm), which is an analytical upper bound value that is intended to include a) the CR normal operation intake flow rate (including test measurements uncertainties), and b) all unfiltered air inleakage, including a 10 cfm allowance for ingress and egress.

The licensee states that the CR emergency filtered ventilation intake flow varies between 800 to 1,000 cfm (from 600 to 1,030 cfm), which includes allowance for measurement uncertainties.

The licensee states that it used the minimum intake flow rate of 800 cfm in the pressurized mode because although the filtered intake of radioisotopes is higher at the larger intake rate of 1,000 cfm, it is small compared to the radioactivity entering the CR in both cases due to unfiltered air inleakage. Consequently, the depletion of airborne activity in the CR via the higher outleakage rate of 1,000 cfm makes the lower intake rate of 800 cfm more limiting from a dose consequence perspective. The staff confirmed that the dose model using the minimum intake flow rate of 800 cfm in the pressurized mode is more conservative than 1,000 cfm. Therefore, the staff finds the minimum intake flow rate of 800 cfm in the pressurized mode acceptable.

The licensee states that the unfiltered air inleakage assumed in the dose consequence analyses envelopes the results of recent tracer gas testing, includes 10 cfm for ingress and egress, and provides margin for potential deterioration between surveillance tests. The licensee also states

that the maximum unfiltered air inleakage into the CRE during the listed modes of operation is assumed to be as follows:

CR Isolation mode - Increased from a previous value of 300 cfm to 450 cfm (this represents an upper bound analytical value which includes test measurement uncertainties and a 10 cfm allowance for ingress and egress).

CR Emergency mode - Increased from a previous value of 30 cfm to 165 cfm (this represents an upper bound analytical value which includes test measurement uncertainties and a 10 cfm allowance for ingress and egress).

The following lists the analysis assumptions and key parameter values that were proposed for the Beaver Valley common CR. Based on calculated control room doses that are within the regulatory limits as provided in the application, the staff review finds that these changes are acceptable.

Control Room Parameters Current Value Previous Value Normal Operation Unfiltered Air Intake 1250 cfm 500 cfm (includes unfiltered air inleakage and a 10 cfm allowance for ingress/egress)

Isolation Mode Unfiltered Air Inleakage 450 cfm 300 cfm (includes 10 cfm for ingress/egress)

Emergency Mode Filtered Air Intake 800 to 1000 cfm 600 to 1030 cfm Emergency Mode Unfiltered Air 165 cfm 30 cfm Inleakage (includes 10 cfm for ingress/egress) 3.1.10 Control Room Habitability and Control Room Envelope Unfiltered Air Inleakage Test Extension As stated in the LAR, TS 5.5.14 requires, in part, that the unfiltered air inleakage past the CRE boundary into the CRE be tested at the frequency specified in Section C.1 of RG 1.197, which states, in part, that CREs should be tested on a performance-based frequency consistent with Figure 1. Per Figure 1, if a CRE test fails and the test acceptance criterion is increased through recalculation of the consequences to the CR operators, a retest is not required. However, a full test should be conducted 3 years later to ascertain whether the CREs integrity has degraded.

In Section 3.4 of the LAR, the licensee described the reason for the request for the original version of the proposed note in TS 5.5.14.c. The licensee supplemented the original request in its letter dated April 14, 2020. The supplement contained the licensees final version, explanation, and justification for the note in TS 5.5.14.c, which creates the one-time extension of the unfiltered air inleakage test. The licensee provided the following explanation regarding the integrity of the boundary and why a test extension is appropriate:

Energy Harbor Nuclear Corp. does not expect the CRE pressurization (emergency) mode boundary to continue to degrade. The CRE boundary will continue to be monitored for potential degradation through normal periodic assessments, maintained in accordance with site preventative maintenance programs, and verified to be acceptable through testing including the following:

flow and filter efficiency tests for Unit 1 and Unit 2 pressurization fan subsystems; differential pressure readings between CRE areas and adjacent non-CRE areas in all modes of operation; and purge fan flow verifications.

A test extension is appropriate for two reasons:

1) The CRE pressurization (emergency) mode unfiltered air inleakage has remained consistent over the most recent three years of testing (2015 to 2017).

2) In the unlikely event of CRE boundary degradation, extending the test interval is not expected to significantly reduce a margin of safety because the proposed amendment would provide unfiltered air inleakage test margin compared to the most recent test results.

As stated in the staffs SE for Units 1 and 2, Amendment Nos. 281 and 163, respectively, one of the reasons for the original testing frequency regimen in TS 5.5.14.c was that the regimen provided assurance that significant degradation of the CRE boundary will not occur between CRE inleakage determinations. The staff reviewed the licensees final version of the proposed note for TS 5.5.14.c, as well as the explanation and justification for the note. The staff concluded that, for the one-time extension period, there is reasonable assurance that significant degradation of the CRE boundary will not occur between CRE inleakage determinations. The staff determined that TS 5.5.14.c, as amended by the proposed note, would continue to meet the requirements of 10 CFR 50.36(c)(5) because TS 5.5.14 would still assure operation of the facility in a safe manner. Therefore, the NRC staff finds the proposed change to TS 5.5.14.c acceptable.

The licensee is proposing to extend the full test period to the normal 6-year frequency instead of the 3-year retest frequency and states that it has high confidence that the CRE has not continued to degrade based upon improving test results between 2015 and 2017, due to plant modifications and system rebalancing efforts for the normal and recirculation (isolation) modes.

The applicant states that for the Units 1 and 2 pressurization (emergency) modes, the step change that occurred between 2008 and 2015 has remained essentially unchanged over the most recent 2-year test period; therefore, it is not anticipated to degrade any further, and the proposed acceptance criteria in this LAR are expected to provide enough margin for potential CRE boundary degradation in the future.

The licensee provided the historical test data in cfm, which included the uncertainty and 10 cfm for ingress and egress as required per RG 1.197 for the CRE tracer gas tests performed at Beaver Valley for each mode of CR ventilation system operation and proposed to change the acceptance criteria for the CRE tracer gas tests as listed below:

Current Acceptance Proposed Acceptance Mode Criteria Criteria Normal 500 cfm 1,250 cfm Recirculation 300 cfm 450 cfm Unit 1 Pressurization 30 cfm 165 cfm Unit 2 Pressurization 30 cfm 165 cfm

The staff finds that the historical test results between 2015 and 2017 confirm that the CRE has not continued to degrade, which the licensee stated was due to plant modifications and system rebalancing efforts for the normal and recirculation (isolation) modes. The staff also finds that the proposed acceptance criteria for all modes of operation bound the results of the latest CRE tracer gas tests performed at Beaver Valley in 2017 with significant margins. Therefore, based on plant modifications and system rebalancing efforts, significant margins to the test acceptance criteria, and calculated control room doses that are within the regulatory limits as provided in the application, the NRC staff review finds that extending the full test period to the normal 6-year frequency instead of the 3-year retest frequency acceptable.

The Unit 1 and 2 CRs are located within a common CRE. The common CRE is served by two ventilation intakes - one for Unit 1 and one for Unit 2. These air intakes are used for both the normal and emergency mode operations. During normal plant operation, both ventilation intakes provide unfiltered outside makeup air. The licensee has proposed to increase the normal operation unfiltered air intake assumed in the dose consequence analyses from the CLB value of 500 cfm to less than or equal to 1,250 cfm. The licensee has proposed to increase the isolation mode unfiltered air inleakage assumed in the dose consequence analyses from the CLB value of 300 cfm to less than or equal to 450 cfm. The licensee has proposed to increase the emergency mode filtered air intake assumed in the dose consequence analyses from the CLB range of 600 cfm to 1,030 cfm to a range of 800 cfm to 1,000 cfm. The licensee has proposed to increase the emergency mode unfiltered inleakage assumed in the dose consequence analyses from the CLB value of 30 cfm to less than or equal to 165 cfm.

For the LOCA, the licensee credits automatic CR isolation 77 seconds after the CIB signal, which is assumed to occur at the start of the accident. For the LOCA, CREA, and MSLB events, the licensee credits the manual activation of the CREVS 30 minutes after accident initiation. For the SGTR, LRA, LACP, FHA, and the SLB, the licensee assumed that the CR is maintained in normal ventilation mode without activating the CREVS during the entire duration of these accidents.

The licensee credits CR purging for a period of 30 minutes at a minimum purge rate of 16,200 cfm commencing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the occurrence of an MSLB accident at either unit. The licensee credits CR purging for a period of 30 minutes at a minimum purge rate of 16,200 cfm commencing 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after the occurrence of an SGTR accident at either unit. The licensee credits CR purging for a period of 30 minutes at a minimum purge rate of 16,200 cfm commencing 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after the occurrence of a Unit 1 FHA. Emergency power is provided to the normal CR ventilation system, including all ventilation system components that are required to support CR operation in the recirculation mode. Credit for CR purging is consistent with the CLB. The NRC staff has already approved the licensee credit for CR purging in several previous amendments, including License Amendment Nos. 275 and 156. Therefore, the NRC staff finds that it is acceptable to credit the normal ventilation system for post-accident CR purging at the times specified in the accident analyses associated with the current LAR.

3.2 Atmospheric Dispersion Estimates The licensee did not propose any changes to the CLB atmospheric dispersion factors for this LAR. The meteorological data, methodology, and previously approved /Q are discussed in the SEs associated with License Amendment Nos. 257 and 139 and License Amendment No. 273 for Unit 1. For completeness, Tables 2A through 2H in Section 5 of this SE are included to indicate the specific atmospheric dispersion factors used in the CR dose consequence analyses

for the eight accidents evaluated by the licensee. Table 3 in Section 5 of this SE shows the offsite atmospheric dispersion factors used for all the licensees dose consequence analyses.

4.0 TECHNICAL CONCLUSION As described above, the NRC staff reviewed the assumptions, inputs, and methods used by the licensee to assess the radiological consequences of DBAs at Beaver Valley, Units 1 and 2. The NRC staff finds that the licensee used assumptions, inputs, and methods consistent with the regulatory requirements and guidance identified in Section 2.0 of this SE. The NRC staff compared the doses estimated by the licensee to the applicable criteria identified in Section 2.0 of this SE. Therefore, the NRC staff concludes, with reasonable assurance, that the licensees estimates of the EAB, LPZ, and CR doses will comport with the licensees estimates. The NRC staff further finds reasonable assurance that Beaver Valley, Units 1 and 2, as modified by these license amendments, will continue to provide sufficient safety margins with adequate defense in depth to address unanticipated events and to compensate for uncertainties in accident progression and analysis assumptions and parameters.

Based on the staffs review of the licensees analyses discussed above, the proposed TS changes are justified and acceptable. The staff determined that 10 CFR 50.36(b) will continue to be met because the TSs will continue to be based on the analyses and evaluations provided by the UFSARs and amendments thereto. The staff determined that the remedial actions of TS 3.4.16, as amended by the proposed change, are acceptable because they will continue allow safe operation of the facility in the event the LCO is not met. The staff determined the LCO statement of TS 3.7.14, as amended by the proposed changes, will continue to require the lowest functional capability or performance levels of equipment required for safe operation of the facility. Therefore, the staff determined that the requirements of 10 CFR 50.36(c)(2) will continue to be met. The staff determined that SRs 3.4.16.2 and 3.7.13.1, as amended by the proposed changes, will continue to assure that the necessary quality of systems and components is maintained and that the LCOs will be met. Therefore the requirements of 10 CFR 50.36(c)(3) will continue to be met. The staff determined that TS 5.5.14.c, as amended by the proposed note, would continue to meet the requirements of 10 CFR 50.36(c)(5) because it would still assure operation of the facility in a safe manner.

The TSs, as amended by the changes, will continue to be based on the analyses and evaluations provided by the UFSARs and amendments thereto, and 10 CFR 50.36 will continue to be met. Therefore the NRC staff finds the proposed changes to TSs 3.4.16, 3.7.13, 5.5.7, and 5.5.14 acceptable.

5.0 TABLES The results of the evaluations performed by the licensee, as well as the applicable dose acceptance criteria from RG 1.183, are shown in Table 1 in this section of the SE. CR atmospheric dispersion factors are shown in Tables 2A through 2H in this section of the SE.

Offsite atmospheric dispersion factors are shown in Table 3, and CR data and assumptions are shown in Table 4 in this section of the SE. Accident-specific data and assumptions are shown in Tables 5 through 11 in this section of the SE. Finally, the assumptions found acceptable to the NRC staff are presented in Table 12 in this section of the SE. These tables are referenced in Section 3 of this SE.

Table 1 Beaver Valley, Units 1 and 2, Radiological Consequences Expressed as TEDE (1) (rem)

Design-Basis Accidents EAB (2) LPZ (3) CR Loss-of-Coolant Accident 16.62 2.9 4.5 Main Steamline Break Accident (Unit 1) (4) 0.11 0.02 1.7 Main Steamline Break Accident (Unit 2) (4) 0.4 0.1 1.5 Steam Generator Tube Rupture(Unit 1) (4) 2.3 0.14 2.6 Steam Generator Tube Rupture(Unit 2) (4) 1.3 0.08 0.4 Dose Acceptance Criteria 25 25 5 Control Rod Ejection Accident (6) 3 1.4 3.1 Control Rod Ejection Accident (7) 1 0.1 0.2 Fuel Handling Accident (Unit 1) 2.1 0.12 4.2 Fuel Handling Accident (Unit 2) 2.5 0.12 1.4 Dose Acceptance Criteria 6.3 6.3 5 Main Steamline Break Accident (Unit 1)(5) 0.14 0.04 1.7 Main Steamline Break Accident (Unit 2)(5) 2.5 0.7 1.5 Steam Generator Tube Rupture (Unit 1)(5) 0.88 0.06 2.6 Steam Generator Tube Rupture (Unit 2)(5) 0.64 0.05 0.4 Locked Rotor Accident (Unit 2)(8) 2.3 0.35 2.9 Loss of AC Power Note 9 Note 9 Note 9 Small Line Break Outside Containment 0.22 0.011 0.7 Dose Acceptance Criteria 2.5 2.5 5 (1) Total effective dose equivalent (2) Exclusion area boundary (3) Low population zone (4) Pre-accident iodine spike (5) Concurrent iodine spike (6) Assumes containment release (7) Assumes secondary side release (8) For Unit 1, the dose from a postulated locked rotor accident is bounded by the Unit 2 locked rotor accident.

(9) The dose from a postulated loss of alternating current power is bounded by the locked rotor accident.

Table 2A Beaver Valley, Units 1 and 2, LOCA CR and ERF Atmospheric Dispersion Factors (sec/m3)

Release Location 0 to 2 hr 2 to 8 hr 8 to 24 hr 24 to 96 hr 96 to 720 hr CR Bounding Dispersion Factors Containment leakage 8.16 x 10-4 5.78 x 10-4 2.53 x 10-4 2.00 x 10-4 1.78 x 10-4 ESF leakage 8.16 x 10-4 5.78 x 10-4 2.27 x 10-4 1.71 x 10-4 1.47 x 10-4 RWST leakage 7.34 x 10-4 6.17 x 10-4 2.54 x 10-4 1.96 x 10-4 1.57 x 10-4

ERF Bounding Dispersion Factors Containment leakage 7.22 x 10-5 6.43 x 10-5 2.96 x 10-5 2.48 x 10-5 2.15 x 10-5 ESF leakage 7.22 x 10-5 6.43 x 10-5 2.96 x 10-5 2.48 x 10-5 2.15 x 10-5 RWST leakage 9.42 x 10-5 8.37 x 10-5 3.81 x 10-5 2.97 x 10-5 2.58 x 10-5 Table 2B Beaver Valley, Units 1 and 2, CREA CR Atmospheric Dispersion Factors (sec/m3)

Release Location 0 to 2 hr 2 to 8 hr 8 to 24 hr 24 to 96 hr 96 to 720 hr Containment leakage 8.16 x 10-4 5.78 x 10-4 2.53 x 10-4 2.00 x 10-4 1.78 x 10-4 MSSVs/ADVs 1.24 x 10-3 9.94 x 10-4 4.08 x 10-4 3.03 x 10-4 2.51 x 10-4 Table 2C Beaver Valley, Units 1 and 2, MSLB CR Atmospheric Dispersion Factors (sec/m3)

(The dashed lines indicate that releases were terminated prior to the specified time period)

Release Location 0 to 2 hr 2 to 8 hr 8 to 24 hr 24 to 96 hr 96 to 720 hr Unit 1 MS Line break 1.05 x 10-2 7.72 x 10-3 3.01 x 10-3 ------ ------

location Unit 1 MSSVs/ADVs 1.24 x 10-3 9.94 x 10-4 ------ ------ ------

Unit 2 MS Line break 1.03 x 10-3 7.84 x 10-4 3.57 x 10-4 ------ ------

location Unit 2 MSSVs/ADVs 5.01 x 10-4 3.58 x 10-4 ------ ------ ------

Table 2D Beaver Valley, Units 1 and 2, SGTR CR Atmospheric Dispersion Factors (sec/m3)

(The dashed lines indicate that releases were terminated prior to the specified time period)

Release Location 0 to 2 hr 2 to 8 hr 8 to 24 hr 24 to 96 hr 96 to 720 hr Unit 1 Condenser Air Ejector 1.05 x 10-2 ------ ------ ------ ------

(TB SE Corner)

Unit 1 MSSVs/ADVs 1.24 x 10-3 9.94 x 10-4 4.08 x 10-4 ------ ------

Unit 2 Condenser Air Ejector 1.03 x 10-3 ------ ------ ------ ------

(TB NW Corner)

Unit 2 MSSVs/ADVs 5.01 x 10-4 3.58 x 10-4 ------ ------ ------

Table 2E Beaver Valley, Units 1 and 2, LRA CR Atmospheric Dispersion Factors (sec/m3)

(The dashed lines indicate that releases were terminated prior to the specified time period) 96 to 720 Release Location 0 to 2 hr 2 to 8 hr 8 to 24 hr 24 to 96 hr hr Secondary System leakage 1.24 x 10-3 9.94 x 10-4 ------ ------ ------

Table 2F Beaver Valley, Units 1 and 2, Loss of AC Power CR Atmospheric Dispersion Factors (sec/m3)

(The dashed lines indicate that releases were terminated prior to the specified time period)

Release Location 0 to 2 hr 2 to 8 hr 8 to 24 hr 24 to 96 hr 96 to 720 hr Secondary System leakage 1.24 x 10-3 9.94 x 10-4 ------ ------ ------

Table 2G Beaver Valley, Units 1 and 2, FHA CR Atmospheric Dispersion Factors (sec/m3)

Release Location 0 to 2 hr 2 to 8 hr 8 to 24 hr 24 to 96 hr 96 to 720 hr CR Bounding Dispersion Factors (Unit 1)

Unit 1 Ventilation Vent 4.75 x 10-3 3.66 x 10-3 1.43 x 10-3 1.02 x 10-3 8.84 x 10-4 CR Bounding Dispersion Factors (Unit 2)

Unit 2 Ventilation Vent 9.39 x 10-4 6.69 x 10-4 3.08 x 10-4 2.23 x 10-4 1.54 x 10-4 Table 2H Beaver Valley, Units 1 and 2, Small Line Break Outside Containment CR Atmospheric Dispersion Factors (sec/m3)

(The dashed lines indicate that releases were terminated prior to the specified time periods)

Release Location/Receptor 0 to 2 hr 2 to 8 hr 8 to 24 hr 24 to 96 hr 96 to 720 hr Unit No. 1 Ventilation Vent 4.75 x 10-3 ------ ------ ------ ------

Table 3 Beaver Valley, Units 1 and 2, Offsite Atmospheric Dispersion Factors (sec/m3)

Exclusion Area Boundary Release Location 0 to 2 hr Unit No. 1 Release Points 1.04 x 10-3 Unit 2 Release Points 1.25 x 10-3 Low Population Zone Release Location 0 to 8 hr 8 to 24 hr 24 to 96 hr 96 to 720 hr Unit No. 1 & Unit 2 Release 6.04 x 10-5 4.33 x 10-5 2.10 x 10-5 7.44 x 10-6 Points

Table 4 Beaver Valley, Units 1 and 2, CR Data Assumptions CR envelope minimum free air volume 173,000 ft3 CR purge flowrate and time - minimum 16,200 cfm for 30 min Normal operational maximum intake/inleakage 1,250 cfm Isolation mode unfiltered inleakage with 10 cfm ingress/egress 450 cfm Emergency mode filtered air intake 800 to 1,000 cfm Emergency mode filtered recirculation Not Credited Emergency mode intake filter efficiency Elemental iodine 98%

Organic iodine 98%

Particulates 99%

Emergency mode filtered recirculation Not Credited Emergency mode maximum unfiltered inleakage with 10 cfm 165 cfm CR purge flowrate and time - minimum 16,200 cfm for 30 min Operator breathing rate 0 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 3.5 x10-4 m3/sec Occupancy factors 0 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.0 24 - 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 0.6 96 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 0.4 Delay in CREVS initiation with assumed loss of off-site power Automatic on CIB signal Automatic isolation EDG startup and T = 77 seconds load sequencing Emergency mode automatic initiation T = 137 seconds Not credited in dose analysis Manual Emergency mode after auto isolation T = 30 minutes

Table 5 (Page 1 of 2)

Beaver Valley, Units 1 and 2, Data and Assumptions for the LOCA Containment Leakage Power level for radiological source term 2,918 MWt Minimum containment free volume 1,750,000 ft3 Containment leak rate (0-24 hr) 0.1% volume fractions per day Containment leak rate (1-30 day) 0.05% volume fractions per day Containment spray coverage 60%

Containment spray period 77.4 seconds to 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months Containment mixing rate sprayed/unsprayed 2 unsprayed volumes per hour Maximum DF for elemental iodine 200 Iodine removal coefficients - sprayed region Technical Report (TR)1 Table 7.2-2 Iodine removal coefficients - unsprayed region TR1 Table 7.2-2 Sump/recirculation spray pH 7 Iodine chemical form in containment atmosphere 4.85% elemental 95% particulate 0.15% organic Release point Worst case between containment Wall & SLCRS vent (containment ESF Leakage Minimum sump volume (5 to 30 min) 19,253 ft3 (1.1379 x 106 lbm)

(30 min to 2 hr) 24,909 ft3 (1.5133 x 106 lbm)

(2 hr to 30 day) 43,824 ft3 (2.6837 x 106 lbm)

ESF expected leak rate 5,700 cc/hr ESF leak rate used in analysis 11,400 cc/hr (2 x expected)

Duration of ESF leakage 1,200 seconds to 30 days ESF iodine release fraction 0.1 Chemical form of iodine released 97% elemental; 3% organic ESF unfiltered PUFF release point SLCRS vent (containment top)

RWST (Back-Leakage)

Sump water back-flow to RWST 1 gpm Sump water back-flow to RWST used in analysis 2 gpm Onset of back-leakage 1,768 seconds Onset of RWST activity venting 3,039 seconds End of release period 30 days Iodine release fraction via RWST TR1 Figure 7.2-3 and Table 7.2-3 RWST release point RWST vent 1Technical Report 101173-RADR-002-01 dated August 12, 2019, included in the April 14, 2020, supplement to the LAR.

Table 5 (Page 2 of 2)

Beaver Valley, Units 1 and 2, Data and Assumptions for the LOCA Containment Vacuum System Reactor coolant TS activity TR1 Table 4.2-1A Chemical form of iodine released 97% elemental; 3% organic Containment vacuum system release 2,200 scfm for 5 sec Release point Worst case between Containment Wall

& SLCRS Vent (Containment Top)

CREVS Initiation Signal Timing Initiation time (signal) Containment Isolation Phase B (CIB) signal assumed to actuate at t = 0 CR isolation time t = 77 seconds (automatic)

Emergency mode initiation time t = 30 minutes (manual initiation) 1Technical Report 101173-RADR-002-01 dated August 12, 2019, included in the April 14, 2020, supplement to the LAR.

Table 6 Beaver Valley, Units 1 and 2, Data and Assumptions for the CRE Accident Containment Pathway Parameters Power level for radiological source term 2,918 MWt Containment minimum free volume 1.75 x 106 ft3 Containment leak rate (0-24 hr) 0.1% volume fractions per day Containment leak rate (1-30 day) 0.05% volume fractions per day Failed fuel percentage <10%

Percentage of core inventory in fuel gap 10% (noble gases and halogens)

Melted fuel percentage < 0.25%

Percentage of core inventory in melted fuel released 100% noble gas; 25% halogens to containment atmosphere Chemical form of iodine in failed/melted fuel 4.85% elemental; 95% CsI 0.15% organic Radial peaking factor 1.70 Core activity release timing Instantaneous (Puff)

Form of iodine in the containment atmosphere 97% elemental; 3% organic Termination of containment release 30 days Environmental release point Containment wall/SLCRS vent (containment top)

Secondary Side Pathway Parameters Minimum reactor coolant mass 341,331 lbm Primary-to-secondary leakage 150 gpd per SG at STP, 450 gpd total Termination of primary-to-secondary leakage 2,500 secs Fraction of failed/melted fuel Same as Containment pathway

Percentage of core inventory in melted fuel 100% noble gas; 50% halogens released to reactor coolant Iodine species released to environment 97% elemental; 3% organic Iodine partition coefficient 100 (all tubes submerged)

Fraction of noble gas released 1.0 (released without holdup)

Minimum post-accident SG liquid mass 99,217 lbm per SG Steam releases per SG 0 to 150 secs 900 lbs/sec 150 to 300 secs 300 lbs/sec 300 to 2500 secs 150 lbs/sec 2500 secs to 8 hrs 778,000 lbs 8 hrs to 30 days Rate of release conservatively assumed to be same as that applicable to the previous time period.

Termination of Release from SGs 30 days Environmental Release Point MSSVs/ADVs CREVS Initiation Signal Timing Emergency mode initiation time t = 30 minutes (manual initiation)

Table 7A Beaver Valley, Units 1, Data and Assumptions for the MSLB Accident Source Term Parameters Core Power Level 2,918 MWt Minimum Reactor Coolant Mass 345,097 lbm Leakrate into Faulted SG 150 gpd at STP Amount of Accident-Induced Leakage into Faulted SG N/A Maximum Time to Cool RCS to 212 ºF 19 hrs Leakrate into Intact SGs 300 gpd total from 2 SGs at STP RCS TS Iodine and Noble Gas Activity Concentration TR1 Table 4.2-1A (0.35 Ci/gm Reactor Coolant Equilibrium Iodine Appearance Rates TR1 Table 4.2-2A (0.35 Ci/gm Pre-Accident Iodine Spike Activity TR1 Table 4.2-2A (21 Ci/gm Concurrent Iodine Spike Appearance Rate 500 Xs equilibrium appearance Duration of Concurrent Iodine Spike 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Secondary System Release Parameters Iodine Species released to Environment 97% elemental; 3% organic Secondary Coolant TS Iodine Activity Concentration TR1 Table 4.2-1A (0.10 Ci/gm Iodine Partition Coefficient in Intact SG 100 (all tubes submerged)

Fraction of Noble Gas Released from Intact SG 1.0 (released without holdup)

Fraction of Iodine Released form Faulted SG 1.0 (released without holdup)

Fraction of Noble Gas Released from Faulted SG 1.0 (released without holdup)

Initial and Minimum Post-Accident Intact SG Liquid Mass 101,799 lbm per SG

Maximum Liquid in Faulted SG 163,150 lbm Steam Releases from Intact SG 345,000 lbm (0 to 2 hr) 734,000 lbm (2 to 8 hr)

Dryout of Faulted SG Instantaneous (Puff release)

Termination of Release from Faulted SG 19 hours2.199074e-4 days 0.00528 hours 3.141534e-5 weeks 7.2295e-6 months Termination of Release from Intact SG 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Environmental Release Point: Faulted SG Break Point Environmental Release Point: Intact SG MSSVs/ADVs CREVS Initiation Signal Timing Emergency Mode Initiation Time t = 30 minutes (manual initiation)

CR Purge (Time/Rate) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the MSLB 30 minutes at 16,200 cfm 1Technical Report 101173-RADR-002-01 dated August 12, 2019, included in the April 14, 2020, supplement to the LAR.

Table 7B Beaver Valley, Unit 2, Data and Assumptions for the MSLB Accident Source Term Parameters Core Power Level 2,918 MWt Minimum Reactor Coolant Mass 341,331 lbm Leakrate into Faulted Steam Generator 150 gpd at STP Amount of Accident-Induced Leakage into Faulted SG 8.1 gpm at STP Maximum Time to Cool RCS to 212 ºF 21 hrs Leakrate into Intact Steam Generators 300 gpd total from 2 SGs at STP RCS TS Iodine and Noble Gas Activity Concentration TR1 Table 4.2-1B (0.10 Ci/gm Reactor Coolant Equilibrium Iodine Appearance Rates TR1 Table 4.2-2B (0.10 Ci/gm Pre-Accident Iodine Spike Activity TR1 Table 4.2-2B (6 Ci/gm DEI)

Concurrent Iodine Spike Appearance Rate 500 Xs equilibrium appearance Duration of Concurrent Iodine Spike 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Secondary System Release Parameters Iodine Species Released to Environment 97% elemental; 3% organic Secondary Coolant TS Iodine Activity Concentration TR1 Table 4.2-1B (0.05 Ci/gm Iodine Partition Coefficient in Intact SG 100 (all tubes submerged)

Fraction of Noble Gas Released from Intact SG 1.0 (released without holdup)

Fraction of Iodine Released from Faulted SG 1.0 (released without holdup)

Fraction of Noble Gas Released from Faulted SG 1.0 (released without holdup)

Initial & Minimum Post-Accident Intact SG Liquid Mass 105,076 lbm per SG Maximum Liquid in Faulted SG 162,800 lbm Steam Releases from Intact SG 350,000 lbm (0 to 2 hr) 730,000 lbm (2 to 8 hr)

Dryout of Faulted SG Instantaneous (Puff release)

Termination of Release from Faulted SG 21 hours2.430556e-4 days 0.00583 hours 3.472222e-5 weeks 7.9905e-6 months Termination of Release from Intact SG 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Environmental Release Point: Faulted SG Break Point Environmental Release Point: Intact SG MSSVs/ADVs CREVS Initiation Signal Timing Emergency Mode Initiation Time t = 30 minutes (manual initiation)

CR Purge (Time/Rate) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the MSLB 30 minutes at 16,200 cfm 1Technical Report 101173-RADR-002-01 dated August 12, 2019, included in the April 14, 2020, supplement to the LAR.

Table 8A Beaver Valley, Unit 1, Data and Assumptions for the SGTR Accident Source Term Parameters Core Power Level 2,918 MWt Minimum Reactor Coolant Mass 373,100 lbm Break Flow to Ruptured Steam Generator 21,900 lbm (0 to 225 sec) 128,000 lbm (225 to 1800 sec)

Time of Reactor Trip 225 sec Termination of Release from Ruptured SG 1,800 seconds Fraction of Break Flow that Flashes 0.2227 (0 to 225 sec) 0.1645 (225 to 1800 sec)

Leakage Rate to Intact Steam Generators 150 gpd at STP for each SG Failed/Melted Fuel Percentage 0%

RCS TS Iodine and Noble Gas Activity Concentration TR1 Table 4.2-1A (0.35 Ci/gm DEI)

Reactor Coolant Equilibrium Iodine Appearance Rates TR1 Table 4.2-2A (0.35 Ci/gm DEI)

Pre-Accident Iodine Spike Activity TR1 Table 4.2-2A (21 Ci/gm DEI)

Concurrent Iodine Spike Appearance Rate 335 Xs equilibrium appearance rates Duration of Concurrent Iodine Spike 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Secondary System Release Parameters Intact SG Liquid Mass (minimum) 91,000 lbm Ruptured SG Liquid Mass (minimum) 91,000 lbm Initial SG Liquid Mass per SGs 96,000 lbm Secondary Coolant TS Iodine Activity Concentration Table 4.2-1A (0.10 Ci/gm DE I-131)

Iodine Species Released to Environment 97% elemental; 3% organic Iodine Partition Coefficient (unflashed portion) 100 (all tubes submerged)

Fraction of Iodine Released (flashed portion) 1.0 (released without holdup)

Fraction of Noble Gas Released from any SG 1.0 (released without holdup)

Partition Factor in Condenser 100 elemental iodine 1 organic iodine/noble gases

Steam Flowrate to Condenser Before Reactor Trip 1,207.407 lbm/sec per SG (0 to 225 sec)

Ruptured SG Steam Releases via MSSVs/ADVs 68,900 lbm (225 to 1800 sec)

Intact SG Steam Releases via MSSVs/ADVs 417,100 lbm (225 to 7200 sec) 979,500 lbm (2 to 8 hr) 658,400 lbm (8 to 16 hr) 546,700 lbm (16 to 24 hr)

Termination of Release from Intact SGs 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Environmental Release Points Condenser Air Ejector (0 to 225 sec)

MSSVs/ADVs (225 sec to 24 hr)

CREVS Initiation Signal Timing CR Ventilation is Maintained in Normal Mode T= 0 to T= 8 Hours CR Manual Purge Initiation at T = 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 30 min purge at 16,200 cfm 1Technical Report 101173-RADR-002-01 dated August 12, 2019, included in the April 14, 2020, supplement to the LAR.

Table 8B Beaver Valley, Unit 2, Data and Assumptions for the SGTR Accident Source Term Parameters Core Power Level 2,918 MWt Minimum Reactor Coolant Mass 368,000 lbm Break Flow to Ruptured Steam Generator 9,200 lbm (0 to 116 sec) 197,400 lbm (116 to 4076 sec)

Time of Reactor Trip 116 seconds Termination of Release from Ruptured SG 4,076 seconds Amount of Break Flow that Flashes 1,730.2 lbm (0 to 116 sec) 6,814.5 lbm (116 to 1932.5 sec)

Leakage Rate to Intact Steam Generators 150 gpd at STP for each SG Failed/Melted Fuel Percentage 0%

RCS TS Iodine and Noble Gas Activity Concentration TR1 Table 4.2-1A (0.35 Ci/gm DEI)

Reactor Coolant Equilibrium Iodine Appearance Rates TR1 Table 4.2-2A (0.35 Ci/gm DEI)

Pre-Accident Iodine Spike Activity TR1 Table 4.2-2A (21 Ci/gm DEI)

Concurrent Iodine Spike Appearance Rate 335 Xs equilibrium appearance rates Duration of Concurrent Iodine Spike 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Secondary System Release Parameters Intact SG Liquid Mass (minimum) 95,150 lbm Ruptured SG Liquid Mass (minimum) 95,150 lbm Initial SG Liquid Mass per Steam Generators TR1 Table 4.2-1A (0.10 Ci/gm DEI)

Iodine Species released to Environment 97% elemental; 3% organic Iodine Partition Coefficient (unflashed portion) 100 (all tubes submerged)

Fraction of Iodine Released (flashed portion) 1.0 (released without holdup)

Fraction of Noble Gas Released from any SG 1.0 (released without holdup)

Partition Factor in Condenser 100 for elemental iodine only Steam Flowrate to Condenser before Reactor Trip 142,300 lbm (ruptured SG) 281,900 lbm (intact SGs)

Ruptured SG Steam Releases via MSSVs/ADVs 67,300 lbm (116 to 4,076 sec) 0 lbm (4,076 to 7,200 sec) 46,800 lbm (7,200 to 28,800 sec)

Intact SG Steam Releases via MSSVs/ADVs 163,500 lbm (116 to 4,076 sec) 216,800 lbm (4,076 to 7,200 sec) 798,500 lbm (7,200 to 28,800 sec)

Termination of Release from Intact SGs 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Environmental Release Points Condenser air ejector (0 to 116 sec)

MSSVs/ADVs (116 sec to 8 hr)

CREVS Initiation Signal Timing CR ventilation is maintained in Normal mode T= 0 to T= 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months CR Manual Purge initiation at T = 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 30 min purge at 16,200 cfm 1Technical Report 101173-RADR-002-01 dated August 12, 2019, included in the April 14, 2020, supplement to the LAR. The licensee used Unit 1 TS values for coolant concentrations for added conservatism in the Unit 2 SGTR analysis.

Table 9 Beaver Valley, Units 1and 2, Data and Assumptions for the Locked Rotor Accident Source Term Parameters Core power level 2,918 MWt Minimum Reactor Coolant Mass 341,331 lbm Primary-to-Secondary SG tube leakage 450 gpd at STP Failed Fuel Cladding Percentage < 20%

Melted Fuel Percentage 0%

Radial Peaking Factor 1.70 Fraction of Core Inventory in Fuel Gap I-131 (8%)

I-132 (23%)

Kr-85 (35%)

Other Noble Gases (4%)

Other Halides (5%)

Alkali Metals (46%)

Core Activity of Isotopes in Gap TR1 Table 4.3-1 Iodine Chemical Form in Gap 4.85% elemental 95% CsI 0.15% organic Secondary Side Parameters Minimum Post-Accident SG Liquid Mass 101,799 lbm per SG Iodine Species released to Environment 97% elemental; 3% organic Iodine Partition Coefficient in SGs 100 (all tubes submerged)

Particulate Carry-Over Fraction in SGs 0.0025 Steam Releases from SGs 348,000 lbm (0 to 2 hr)

778,000 lbm (2 to 8 hr)

Termination of releases from SGs 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Fraction of Noble Gas released to Environment 1.0 (released without holdup)

Environmental Release Point MSSVs/ADVs CREVS Initiation Signal Timing CR ventilation is maintained in Normal mode 1Technical Report 101173-RADR-002-01 dated August 12, 2019, included in the April 14, 2020, supplement to the LAR.

Table 10 Beaver Valley, Units 1 and 2, Data and Assumptions for the Loss of AC Power(1)

Source Term Parameters Core Power Level 2,918 MWt Minimum Reactor Coolant Mass 341,331 lbm Primary-to-Secondary SG tube leakage 450 gpd at STP Failed Fuel Percentage 0%

Melted Fuel Percentage 0%

Reactor Coolant TS Iodine and Noble Gas Activity TR1 Table 4.2-1A (0.35 Ci/gm Secondary Side Parameters Minimum Post-Accident SG Liquid Mass 101,799 lbm per SG Iodine Species released to Environment 97% elemental; 3% organic Secondary Coolant TS Iodine Activity Concentration TR2 Table 4.2-1A (0.10 Ci/gm Iodine Partition Coefficient in SGs 100 (all tubes submerged)

Fraction of Noble Gas Released from SGs 1.0 (released without holdup)

Steam Releases from SGs 348,000 lbm (0 to 2 hr) 778,000 lbm (2 to 8 hr)

Termination of releases from SGs 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Environmental Release Point MSSVs/ADVs CREVS Initiation Signal Timing CR ventilation is maintained in Normal mode 1 Bounding parameter values to encompass an event at either unit.

2 Technical Report 101173-RADR-002-01 dated August 12, 2019, included in the April 14, 2020, supplement to the LAR.

Table 11 Beaver Valley, Units 1 and 2, FHA in Fuel Pool Area or Containment(1)

Data and Assumptions Source Term Parameters Core Power Level 2,918 MWt Number of Rods in Fuel Assemblies 264 Total Number of Fuel Assemblies 157 Number of Damaged Rods 137 Decay Time Prior to Fuel Movement 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months Radial Peaking Factor 1.70 Fraction of Core Inventory in Fuel Gap I-131 (8%)

I-132 (23%)

Kr-85 (35%)

Other Noble Gases (4%)

Other Halides (5%)

Alkali Metals (46%)

Core Activity of Isotopes in Gap with 100 hrs decay TR2 Table 7.7-2 Minimum Depth of Water in Fuel Pool or Reactor 23 ft Fuel Pool and Reactor Cavity Scrubbing DFs Iodine (200)

Noble Gas (1)

Particulates (infinite)

Rate of Release from Fuel into Pool/Cavity Water Instantaneous Duration of Unfiltered Release to the Environment 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Environmental Release Points Plant vent Accident in Fuel Pool Area More Restrictive of Ventilation Vent or SLCRS Vent (Containment Top)

Accident in Containment More Restrictive of Equipment Hatch, Ventilation Vent, Containment Wall, or SLCRS Vent (Containment Top)

CREVS Initiation Signal Timing Unit 1 CR Ventilation is Maintained in Normal Mode for the 2-hour Release Period Followed by a 30-minute CR Purge at 16,200 cfm (Minimum) after 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 16,200 cfm Unit 2 CR Ventilation is Maintained in Normal mode 1Bounding parameter values to encompass an event at either unit.

2Technical Report 101173-RADR-002-01 dated August 12, 2019, included in the April 14, 2020, supplement to the LAR.

Table 12 Beaver Valley, Units 1 and 2, Small Line Break Outside Containment (1)

Data and Assumptions Source Term Parameters Core Power Level 2,918 MWt Minimum Reactor Coolant Mass 341,331 lbm CVCS Letdown Line Break - Mass Flow Rate 16.79 lbm/second Break Flow Flash Fraction 37%

Failed Fuel Cladding Percentage 0%

Melted Fuel Percentage 0%

RCS TS Iodine and Noble Gas Activity Concentration TR2 Table 4.2-1A (0.35 Ci/gm DE I-131)

RCS Equilibrium Iodine Appearance Rates TR2 Table 4.2-2A (0.35 Ci/gm DE I-131)

Concurrent Iodine Spike Appearance Rate 500 times equilibrium appearance rates Duration of Concurrent Iodine Spike 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Iodine Species released to Environment 97% elemental; 3% organic SLCRS Filter Efficiency 0% Not credited Environmental Release Point Ventilation Vent CREVS Initiation Signal Timing CR ventilation is maintained in Normal 1Bounding parameter values to encompass an event at either unit.

2Technical Report 101173-RADR-002-01 dated August 12, 2019, included in the April 14, 2020, supplement to the LAR.

6.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Commonwealth of Pennsylvania official was notified of the proposed issuance of the amendment on August 4, 2020. The Commonwealth official had no comments.

7.0 ENVIRONMENTAL CONSIDERATION

The amendments change SRs or change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the 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 has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (84 FR 68953; December 17, 2019). 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.

8.0 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 amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: J. Parillo R. Clement H. Wagage M. Hamm N. Chien P. Klein Date: September 23, 2020

ML20213A731 *by memorandum **by e-mail OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LA NRR/DRA/ARCB/BC* NRR/DSS/STSB/BC**

NAME VSreenivas LRonewicz KHsueh VCusumano DATE 08/10/2020 08/10/2020 07/26/2020 08/14/2020 OFFICE NRR/DNRL/NCSG/BC** NRR/DSS/SCPB/BC OGC - NLO NRR/DORL/LPL1/BC NAME SBloom BWittick JMcManus JDanna DATE 08/05/2020 07/30/2020 09/04/2020 09/23/2020 OFFICE NRR/DORL/LPL1/PM NAME VSreenivas DATE 09/23/2020

v t e Letter Sequence Approval
EPID:L-2019-LLA-0223, License Amendment Request to Modify Technical Specifications 3.4.16, RCS Specific Activity, 3.7.13 Secondary Specific Activity, 5.5.7, Ventilation Filter Testing Program (Vftp), and 5.5.14, Control Room Envelope. (Approved, Closed)
Initiation Request, Request Acceptance, Acceptance Supplement Administration Questions 12 March 2020 Answers 9 June 2020 Results Approval, Approval Other: ML20288A916
MONTHYEAR2020-03-12 [Table View]

ML20213A731

Text

SUBJECT:

Dear Mr. Penfield:

Enclosures:

Attachment:

Attachment:

1.0 INTRODUCTION

2.1 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

Background

6.0 STATE CONSULTATION

7.0 ENVIRONMENTAL CONSIDERATION

8.0 CONCLUSION

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