ML24031A622
| ML24031A622 | |
| Person / Time | |
|---|---|
| Site: | Vallecitos Nuclear Center |
| Issue date: | 02/28/2024 |
| From: | Geoffrey Wertz NRC/NRR/DANU/UNPL |
| To: | Martinez C GE Hitachi Nuclear Energy |
| References | |
| EPID?L-2023-LLA-0139 | |
| Download: ML24031A622 (108) | |
Text
February 28, 2024 Mr. Carlos Martinez, Site Manager GE Hitachi Nuclear Energy Vallecitos Nuclear Center 6705 Vallecitos Road Sunol, CA 94586
SUBJECT:
GE-HITACHI NUCLEAR ENERGY AMERICAS, LLC - ISSUANCE OF AMENDMENT NO. 26 RE: POSSESSION-ONLY AUTHORIZATION OF THE REACTOR AND FUEL FOR RENEWED FACILITY OPERATING LICENSE NO. R-33 FOR THE NUCLEAR TEST REACTOR (EPID L-2023-LLA-0139)
Dear Mr. Martinez:
The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 26 to Renewed Facility Operating License No. R-33 for the GE-Hitachi Nuclear Energy Americas, LLC Nuclear Test Reactor (NTR) in response to your application dated October 6, 2023 (Agencywide Documents Access and Management System Accession No. ML23279A110), as supplemented by letters dated December 15, 2023 (ML23349A047), and January 26 (ML24026A196),
February 12 (ML24043A044), and February 23, 2024 (ML24054A539). In the letter dated January 26, 2024, the licensee informed the NRC that it had permanently ceased operation of the NTR on December 21, 2023.
The amendment revises Renewed Facility Operating License No. R-33 and the associated technical specifications to remove the authority to operate the NTR, to authorize possession-only of the reactor and fuel, to remove operational requirements not needed for the possession-only status, and to replace the requirement for NRC-licensed reactor operators with certified fuel handlers.
Pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.51, Continuation of license, paragraph (b), [e]ach license for a facility that has permanently ceased operations, continues in effect beyond the expiration date to authorize ownership and possession of the production or utilization facility, until the Commission notifies the licensee in writing that the license is terminated. As such, Renewed Facility Operating License No. R-33 for the NTR shall remain in effect pursuant to 10 CFR 50.51(b).
C. Martinez 2
A copy of the related safety evaluation is also enclosed. If you have any questions, please contact me at (301) 415-0893, or by email at Geoffrey.Wertz@nrc.gov.
Sincerely, Geoffrey A. Wertz, Project Manager Non-Power Production and Utilization Facility Licensing Branch Division of Advanced Reactors and Non-Power Production and Utilization Facilities Office of Nuclear Reactor Regulation Docket No. 50-73 License No. R-33
Enclosures:
1.
Amendment No. 26 to Renewed Facility Operating License No. R-33 2.
Safety Evaluation cc w/enclosures: GovDelivery Subscribers Signed by Wertz, Geoffrey on 02/28/24
ML24031A622 NRR-058 OFFICE NRR/DANU/UNPL/PM NRR/DANU/UNPL/LA OGC/NLO NAME GWertz NParker JWachutka DATE 2/1/2024 2/8/2024 2/26/2024 OFFICE NRR/DANU/UNPL/BC (A)
NMSS/DFM/NARAB/BC NMSS/DUWP/RDB/BC NAME HCruz DJohnson SAnderson DATE 2/26/2024 2/27/2024 2/27/2024 OFFICE NRR/DANU/UNPL/PM NAME GWertz DATE 2/28/2024
GE-HITACHI NUCLEAR ENERGY AMERICAS, LLC DOCKET NO. 50-73 NUCLEAR TEST REACTOR AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 26 License No. R-33
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to Renewed Facility Operating License No. R-33, filed by GE-Hitachi Nuclear Energy Americas, LLC (the licensee) on October 6, 2023, as supplemented by letters dated December 15, 2023, and January 26, February 12, and February 23, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in Title 10 of the Code of Federal Regulations (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 that (i) the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) 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; E.
This issuance of this amendment is in accordance with 10 CFR Part 51, Environmental Protection Regulations for Domestic Licensing and Related Regulatory Functions, of the Commissions regulations, and all applicable requirements have been satisfied; and F.
Prior notice of this amendment was not required by 10 CFR 2.105, Notice of proposed action, and publication of a notice of issuance for this amendment is not required by 10 CFR 2.106, Notice of issuance.
- 2.
Accordingly, the license is amended as described in Attachment 1 to this amendment and by changes to the Technical Specifications as described in Attachment 2. The following paragraphs of Renewed Facility Operating License No. R-33 are hereby amended to read as follows:
2.B.(1)
Pursuant to Section 104c of the Act and 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, to possess the reactor at the designated location in Alameda County, California, in accordance with the procedures and limitations described in the application and set forth in this license.
2.B.(2)
Pursuant to the Act and 10 CFR Part 70, to possess, but not separate:
- a. up to 4 kilograms of contained uranium-235 of an enrichment of 20 percent or greater in the isotope uranium-235, in the form of in-core reactor fuel; and
- b. such special nuclear material as may have been produced by the operation of the reactor.
2.B.(3)
Pursuant to the Act and 10 CFR Part 30, to possess:
- a. up to 0.2-curie radium-beryllium sealed startup source, and up to 0.03 curies of byproduct materials, in the form of sealed sources, for instrument calibration and source checks; and
- b. but not to separate, such byproduct material as may have been produced by the operation of the reactor.
2.C.(1)
Maximum Power Level The licensee is not authorized to operate the reactor at any power level.
2.C.(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised by Amendment No. 26, are hereby incorporated in their entirety in the license.
The licensee shall maintain the facility in accordance with the Technical Specifications.
2.D The license is effective as of the date of its issuance and until the Commission notifies the licensee in writing that the license is terminated.
- 3.
This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Holly D. Cruz, Acting Chief Non-Power Production and Utilization Facility Licensing Branch Division of Advanced Reactors and Non-Power Production and Utilization Facilities Office of Nuclear Reactor Regulation Attachments:
- 1. Changes to Renewed Facility Operating License No. R-33
- 2. Changes to Appendix A, Technical Specifications Date of Issuance: February 28, 2024 Holly D. Cruz Digitally signed by Holly D. Cruz
ATTACHMENT 1 TO LICENSE AMENDMENT NO. 26 RENEWED FACILITY OPERATING LICENSE NO. R-33 DOCKET NO. 50-73 Replace the following pages of Renewed Facility Operating License No. R-33 with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Insert 2
2 3
3 4
4 Amendment No. 26 February 28, 2024 I.
The receipt, possession, and use of byproduct and special nuclear materials as authorized by this license will be in accordance with the Commissions regulations in 10 CFR Part 30, Rules of General Applicability to Domestic Licensing of Byproduct Material, and 10 CFR Part 70, Domestic Licensing of Special Nuclear Material.
- 2. Accordingly, Facility Operating License No. R-33 is hereby renewed in its entirety to read as follows:
A. This license applies to the NTR, which is owned by GE-Hitachi Nuclear Energy Americas, LLC and located at its Vallecitos Nuclear Center in Alameda County, California, and described in the application for license renewal dated November 19, 2020, as supplemented.
B. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses GE-Hitachi Nuclear Energy Americas, LLC:
(1) Pursuant to Section 104c of the Act and 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, to possess the reactor at the designated location in Alameda County, California, in accordance with the procedures and limitations described in the application and set forth in this license.
(2) Pursuant to the Act and 10 CFR Part 70, to possess, but not separate:
- a. up to 4 kilograms of contained uranium-235 of an enrichment of 20 percent or greater in the isotope uranium-235, in the form of in-core reactor fuel; and
- b. such special nuclear material as may have been produced by the operation of the reactor.
(3) Pursuant to the Act and 10 CFR Part 30, to possess:
- a. up to 0.2-curie radium-beryllium sealed startup source, and up to 0.03 curies of byproduct materials, in the form of sealed sources, for instrument calibration and source checks; and
- b. but not to separate, such byproduct material as may have been produced by the operation of the reactor.
Amendment No. 26 February 28, 2024 C. This license shall be deemed to contain and is subject to the conditions specified in 10 CFR Parts 20, 30, 50, 51, 55, 70, and 73 of the Commissions regulations; 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 to the additional conditions specified below:
(1) Maximum Power Level The licensee is not authorized to operate the reactor at any power level.
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised by Amendment No. 26, are hereby incorporated in their entirety in the license. The licensee shall maintain the facility in accordance with the Technical Specifications.
(3) Physical Security Plan The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security plan, including amendments and changes made pursuant to the authority of 10 CFR 50.54, Conditions of licenses, paragraph (p). The approved physical security plan, entitled VNC Site Physical Security Plan, dated April 4, 2022, consists of documents withheld from public disclosure pursuant to 10 CFR 73.21, Protection of Safeguards Information: Performance requirements.
(4) GE-Hitachi Nuclear Energy Americas, LLC, as stated in the General Electric Companys (GEs) application dated January 19, 2007, and supplemented on January 25, 2007, February 23, 2007, March 2, 2007, March 26, 2007, May 16, 2007, May 18, 2007, June 4, 2007, July 6, 2007, and August 9, 2007, will abide by all commitments and representations previously made by GE with respect to the license. These include, but are not limited to, maintaining decommissioning records, implementing decontamination activities, and eventually decommissioning the facility.
(5) The Manager of the Vallecitos Nuclear Center, the Vice-President, Reactor Facility Safety and Security of GE-Hitachi Nuclear Energy Americas, LLC, and the Manager of GE-Hitachi Nuclear Energy Americas, LLC shall be U.S. citizens.
These individuals shall have the responsibility and exclusive authority to ensure and shall ensure that the business and activities of GE-Hitachi Nuclear Energy Americas, LLC, with respect to this license, are at all times conducted in a manner consistent with the protection of the public health and safety and the common defense and security.
Amendment No. 26 February 28, 2024 (6) The commitments/representations made in the GEs application dated January 19, 2007, and supplemented on January 25, 2007, February 23, 2007, March 2, 2007, March 26, 2007, May 16, 2007, May 18, 2007, June 4, 2007, July 6, 2007, and August 9, 2007, regarding reporting relationships and authority over safety and security issues and compliance with NRC requirements shall be adhered to and not be modified without the prior written consent from the Director, Office of Nuclear Reactor Regulation, or designee.
(7) GE-Hitachi Nuclear Energy Americas, LLC shall cause to be transmitted to the Director, Office of Nuclear Reactor Regulation within 30 days of filing with the U.S. Securities and Exchange Commission (SEC), any schedule 13D or 13G filed pursuant to the Securities Exchange Act of 1934 that discloses beneficial ownership of a registered class of General Electric stock.
D. The license is effective as of the date of its issuance and until the Commission notifies the licensee in writing that the license is terminated.
FOR THE NUCLEAR REGULATORY COMMISSION
/original signed by Andrea D. Veil/
Andrea D. Veil, Director Office of Nuclear Reactor Regulation
Attachment:
Appendix A, Technical Specifications Date of Issuance: June 29, 2023
ATTACHMENT 2 TO LICENSE AMENDMENT NO. 26 RENEWED FACILITY OPERATING LICENSE NO. R-33 DOCKET NO. 50-73 Replace all of the pages of Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number.
Remove Insert All All
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Amendment No. 26 February 28, 2024 GE HITACHI NUCLEAR ENERGY NEDO 32765 Revision 8 January 2024 TECHNICAL SPECIFICATIONS FOR THE NUCLEAR TEST REACTOR FACILITY LICENSE R-33 Copyright © 2023, GE-Hitachi Nuclear Energy Americas LLC All Rights Reserved
~ ii ~
Amendment No. 26 February 28, 2024 TABLE OF CONTENTS 1
INTRODUCTION................................................................................................................... 1 1.1 DEFINITIONS................................................................................................................ 1 2
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS.................................... 4 2.1 SAFETY LIMITS............................................................................................................ 4 2.2 LIMITING SAFETY SYSTEM SETTINGS................................................................... 4 3
LIMITING CONDITIONS FOR POLSC (LCP).................................................................... 4 3.1 REACTOR CORE PARAMETERS.............................................................................. 4 3.2 REACTOR CONTROL AND SAFETY SYSTEM........................................................ 4 3.3 PRIMARY COOLANT SYSTEM................................................................................... 5 3.4 CONFINEMENT............................................................................................................ 6 3.5 REACTOR CELL, VENTILATION, AND CONFINEMENT SYSTEM......................... 6 3.6 EMERGENCY POWER................................................................................................ 6 3.7 RADIATION AND ENVIRONMENTAL MONITORING SYSTEMS................................ 6 3.8 EXPERIMENTS............................................................................................................. 7 4
SURVEILLANCE REQUIREMENTS..................................................................................... 7 4.1 REACTOR CORE PARAMETERS................................................................................. 7 4.2 REACTOR CONTROL AND SAFETY SYSTEM............................................................ 8 4.3 PRIMARY COOLANT SYSTEM..................................................................................... 8 4.4 CONFINEMENT............................................................................................................. 9 4.5 REACTOR CELL VENTILATION, AND CONFINEMENT SYSTEM............................... 9 4.6 EMERGENCY POWER.................................................................................................. 9 4.7 RADIATION AND ENVIRONMENTAL MONITORING SYSTEMS................................. 9 4.8 EXPERIMENTS............................................................................................................ 10 5
DESIGN FEATURES........................................................................................................... 10 5.1 SITE AND FACILITY DESCRIPTION........................................................................... 10 5.2 PRIMARY COOLANT SYSTEM................................................................................... 11 5.3 REACTOR CORE AND FUEL...................................................................................... 11 6
ADMINISTRATIVE CONTROLS.......................................................................................... 12 6.1 ORGANIZATION.......................................................................................................... 12 6.2 REVIEW AND AUDIT................................................................................................... 13 6.3 RADIATION SAFETY................................................................................................... 15 6.4 PROCEDURES............................................................................................................ 15 6.5 EXPERIMENTS REVIEW AND APPROVAL................................................................ 16 6.6 REQUIRED ACTIONS.................................................................................................. 16 6.7 REPORTS.................................................................................................................... 17 6.8 RECORDS.................................................................................................................... 18
1 Amendment No. 26 February 28, 2024 1
INTRODUCTION This document constitutes the Technical Specifications for the GEH Nuclear Test Reactor as required by 10 CFR 50.36 and supersedes all prior Technical Specifications. This document includes the basis to support the selection and significance of the specifications. The Technical Specifications are based on the guidance provided in American National Standards Institute/ American Nuclear Society (ANSI/ANS) 15.1-2007, The Development of Technical Specifications for Research Reactors as modified by NUREG-1537, Part 1, Appendix 14.1, Format and Content of Technical Specifications for Non-Power Reactors.
These Technical Specifications provide limits that assure reactor-related activity will be controlled in a way that protects the health and safety of the public, the environment, and on-SITE personnel. Areas addressed are Definitions, Limiting Conditions for POLSC (LCP),
Surveillance Requirements, Design Features, and Administrative Controls.
1.1 DEFINITIONS ADMINISTRATIVE CHANGE(S):
An editorial, non-technical change, which does not affect nuclear safety, personnel safety, security, quality, or change the intent of the document being changed.
CHANNEL(S):
The combination of sensors, lines, amplifiers, and output devices which are connected for the purpose of measuring the value of a parameter.
CHANNEL CALIBRATION:
A comparison and/or an adjustment of the CHANNEL so that its output corresponds with acceptable accuracy to known values of the parameter which the CHANNEL measures.
Calibration SHALL encompass the entire CHANNEL, including equipment actuation, alarm, or trip test and SHALL include the CHANNEL TEST.
CHANNEL CHECK:
A qualitative verification of acceptable performance by observation of CHANNEL behavior. This verification where possible SHALL include comparison of the CHANNEL with other independent CHANNELS or systems measuring the same parameter.
CHANNEL TEST:
The introduction of a signal into the CHANNEL to verify that it is OPERABLE.
2 Amendment No. 26 February 28, 2024 CONFINEMENT:
The enclosure of the overall FACILITY that is designed to limit the release of effluents between the enclosure and its external environment through controlled or defined pathways.
CONTROL ROD(S):
A non-scrammable device having an electric motor drive and containing boron-carbide material.
These rods have been disabled and remain fully inserted and restrained from any movement in the core per the POSSESSION ONLY LICENSE SHUTDOWN CONFIGURATION.
CORE CONFIGURATION:
See POSSESSION ONLY LICENSE SHUTDOWN CONFIGURATION.
EXPERIMENTAL FACILITY or EXPERIMENTAL FACILITIES:
Any location for an experiment which is on or against the external surfaces of the reactor main graphite pack, thermal column, or within any penetration thereof.
FACILITY:
That portion of building 105 composed of the NTR reactor cell, control room, north room, setup room, and south cell.
LICENSE and LICENSEE:
The written authorization (LICENSE R-33), by the responsible authority (The NRC), for an individual or organization to carry out the duties and responsibilities associated with a personnel position, material, or FACILITY requiring licensing.
MANUAL POISON SHEET(S) (MPS):
Manually positioned devices containing cadmium material used to maintain adequate negative reactivity inventory in the reactor to prevent attainment of criticality.
OPERABLE / INOPERABLE:
A system or component is / is not capable of performing its intended function.
OPERATING:
A component or system is performing its intended function.
POSSESSION ONLY LICENSE SHUTDOWN CONFIGURATION (POLSC):
That plant configuration that ensures the reactor will remain subcritical in the credible limiting accident analysis by restraining CONTROL RODS, SAFETY RODS, and MPS in the positions
3 Amendment No. 26 February 28, 2024 assumed in the criticality safety analysis. Details of the NTR POLSC are included in Section 5.
REACTOR SHUTDOWN CONFIGURATION:
All SAFETY RODS, CONTROL RODS, and in-service MANUAL POISON SHEETS SHALL remain in their respective fully inserted positions. See POLSC.
SAFETY ROD(S):
Previously scrammable devices containing boron-carbide material. These devices have been disabled, are fully inserted, and restrained from any movement, and remain in the core per the POLSC.
SHALL, SHOULD, AND MAY:
The word "SHALL" is used to denote a requirement; the word "SHOULD" is used to denote a recommendation; and the word "MAY" is used to denote permission, neither a requirement nor a recommendation.
SITE:
The area within the confines of the Vallecitos Nuclear Center (VNC) controlled by the LICENSEE (Refer to Safety Analysis Report, Figure 2-3.).
SURVEILLANCE INTERVALS:
Biennial - interval not to exceed 30 months.
Annual - interval not to exceed 15 months.
Semi-annual - interval not to exceed 7.5 months.
Quarterly - interval not to exceed 4 months.
Daily - Must be done during the calendar day.
4 Amendment No. 26 February 28, 2024 2
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS These specifications are not applicable due to the reactor being in a POLSC. Reactor operations are not authorized.
2.2 LIMITING SAFETY SYSTEM SETTINGS These specifications are not applicable due to the reactor being in a POLSC. Reactor operations are not authorized.
3 LIMITING CONDITIONS FOR POLSC (LCP) 3.1 REACTOR CORE PARAMETERS Reactor operations are not authorized. Fuel handling in support of defueling is the only activity allowed.
3.2 REACTOR CONTROL AND SAFETY SYSTEM Specification 3.2.1 RODS INOPERABLE All SAFETY RODS and CONTROL RODS SHALL be maintained fully inserted and restrained from any movement per the applicable conditions defined by the POLSC in TS 5.3.1.
3.2.2 MANUAL POISON SHEETS SECURED MPS slots SHALL be maintained per the applicable conditions defined by the POLSC in TS 5.3.1.
Basis Maintaining the FACILITY in accordance with LCP 3.2.1 and 3.2.2 ensures that the reactor remains safely sub-critical with adequate negative reactivity present to ensure reactor criticality does not occur.
5 Amendment No. 26 February 28, 2024 3.3 PRIMARY COOLANT SYSTEM Specification 3.3.1 FORCED FLOW COOLING Forced primary coolant flow SHALL be OPERABLE. If this condition is not met, corrective action SHALL be taken to restore operability within 90 days.
3.3.2 FUEL LOADING TANK FULL The fuel loading tank SHALL be maintained above the low level alarm.
3.3.3 FUEL LOADING TANK LEVEL ALARM The fuel loading tank low level alarm SHALL actuate at 3 feet below the overflow or higher.
A local visible and remote alarm SHALL be available. Corrective action SHALL be taken if this condition is not met.
3.3.4 PRIMARY COOLANT CONDUCTIVITY The specific conductivity of the primary coolant water SHALL be maintained less than 5 µS/cm when averaged over the four most recent quarterly readings.
Corrective action SHALL be taken if this condition is not met.
Basis Maintaining the FACILITY in accordance with LCP 3.3.1, 3.3.2, and 3.3.3 ensures that the primary pump provides for chemical mixing, flow through the primary water cleanup system, and proper operation of the primary conductivity probe. Neither flow through, nor cooling of the fuel is credited in maintaining the fuel subcritical. Fuel loading tank level ensures that the reactor core tank is full and meets the criticality safety assessment bounding initial conditions and that adequate positive pump head pressure exists for primary pump operation. Remote alarm ensures that notification is made that remedial action is needed. The minimum corrosion rate for aluminum in water (< 50°C) occurs at a pH of 6.5. Maintaining water purity below 5 µS/cm based upon an average of quarterly conductivity readings, will maintain the pH between 5.5 and 7.5. Maintaining the FACILITY in accordance with LCP 3.3.4 ensures aluminum corrosion is within acceptable levels.
6 Amendment No. 26 February 28, 2024 3.4 CONFINEMENT This section left intentionally blank.
3.5 REACTOR CELL, VENTILATION, AND CONFINEMENT SYSTEM Specification 3.5.1 REACTOR CELL NEGATIVE PRESSURE The ventilation system SHALL be OPERATING during evolutions that could result in an airborne concentration of one DAC or greater in the reactor cell. A reactor cell negative differential pressure of not less than 0.5 in. of water with respect to the control room SHALL be verified prior to commencing the evolution.
Basis Maintaining the FACILITY in accordance with LCP 3.5.1 ensures that potentially contaminated reactor cell air is released through the ventilation system filters. Since securing the ventilation system would confine airborne radiation in the reactor cell, the purpose of running the ventilation system is to strike a balance between maintaining safe levels for personnel in the FACILITY and minimizing releases to the environment. Therefore, activities that are anticipated to generate an airborne concentration greater than one DAC locally are considered concerning and are performed only when the ventilation system is OPERATING.
As demonstrated in Chapter 13 of the NTR Safety Analysis Report, CONFINEMENT is not required to ensure radiological doses will not exceed 10 CFR 20 allowable limits.
3.6 EMERGENCY POWER This section left intentionally blank.
3.7 RADIATION AND ENVIRONMENTAL MONITORING SYSTEMS Specification 3.7.1 AREA RADIATION MONITOR An operational area radiation monitor* is required in the reactor cell and will alarm at 10 mr/hr or less.
- An operational area radiation monitor SHALL include:
Instrument readout that is visible in the control room.
7 Amendment No. 26 February 28, 2024 a gamma-sensitive instrument.
A local audible alarm.
Alarm indication at a remote monitoring location.
3.7.2 ENVIRONMENTAL MONITORING The VNC SITE utilizes environmental air sampling stations and thermoluminescent (TLD) and/or optically stimulated luminescence (OSL) dosimeters in locations specified by the VNC Environmental Monitoring Manual.
3.7.3 STACK MONITOR OPERABILITY The stack particulate activity monitor SHALL be OPERATING when any evolution is performed in the FACILITY that could generate an airborne concentration greater than one DAC in the reactor cell and will alarm at 2.0E-9 µCi/cc or less. If the monitor is not OPERABLE, all evolutions that could cause such airborne releases SHALL be discontinued within the FACILITY and corrective action taken to restore operability.
Basis The radiation monitoring systems provide information to FACILITY personnel regarding impending or existing danger from excess radiation. The stack particulate activity monitor is placed in service and operated continuously when reactor cell activities are capable of generating an airborne concentration greater than one DAC in the reactor cell. The alarm setpoint is derived from the normal background activity level.
3.8 EXPERIMENTS No specifications are applicable due to the reactor being in a POLSC. Experiments are not authorized, and explosives are no longer stored in the FACILITY.
4 SURVEILLANCE REQUIREMENTS 4.1 REACTOR CORE PARAMETERS No specifications are applicable due to the reactor being in a POLSC. Reactor operations are not authorized. Fuel handling in support of defueling is the only activity allowed.
8 Amendment No. 26 February 28, 2024 4.2 REACTOR CONTROL AND SAFETY SYSTEM Specification 4.2.1 RODS INOPERABLE SAFETY RODS and CONTROL RODS SHALL be verified semi-annually to meet the conditions of the POLSC.
4.2.2 MANUAL POISON SHEETS SECURED MANUAL POISON SHEET covers SHALL be verified semi-annually to be locked in place and the keys removed from the FACILITY.
Basis Surveillance Requirement 4.2.1 ensures that each SAFETY ROD and CONTROL ROD is maintained INOPERABLE as required by the POLSC.
Surveillance Requirement 4.2.2 ensures that each installed MANUAL POISON SHEET remains fully inserted and locked in position as required by the POLSC.
4.3 PRIMARY COOLANT SYSTEM Specification 4.3.1 FORCED FLOW COOLING The primary coolant flow instrument CHANNEL CHECK SHALL be performed quarterly and a CHANNEL CALIBRATION annually.
4.3.2 FUEL LOADING TANK FULL The fuel loading tank level SHALL be visually checked quarterly.
4.3.3 FUEL LOADING TANK LEVEL ALARM The fuel loading tank low level alarm CHANNEL TEST SHALL be performed quarterly.
4.3.4 PRIMARY COOLANT CONDUCTIVITY The primary coolant conductivity instrument CHANNEL CHECK SHALL be performed quarterly and a CHANNEL CALIBRATION biennially.
9 Amendment No. 26 February 28, 2024 Basis Surveillance Requirement 4.3.1, 4.3.2, and 4.3.3 ensure that primary coolant flow can be initiated and monitored allowing the primary cleanup system to operate efficiently.
Surveillance Requirement 4.3.4 ensures that primary coolant conductivity can be accurately monitored.
4.4 CONFINEMENT This section left intentionally blank.
4.5 REACTOR CELL VENTILATION, AND CONFINEMENT SYSTEM Specifications 4.5.1 REACTOR CELL NEGATIVE PRESSURE The reactor cell differential pressure instrument CHANNEL CHECK SHALL be performed daily when the ventilation system is operating and a CHANNEL CALIBRATION annually.
Basis Maintaining the FACILITY in accordance with Surveillance Requirement 4.5.1 ensures that contaminated reactor cell air is exhausted through the ventilation system. This minimizes the possibility of an airborne contamination release to surrounding areas.
4.6 EMERGENCY POWER This section left intentionally blank.
4.7 RADIATION AND ENVIRONMENTAL MONITORING SYSTEMS Specification 4.7.1 AREA RADIATION MONITOR Area Radiation monitor CHANNEL CHECK SHALL be performed quarterly and a CHANNEL CALIBRATION annually.
10 Amendment No. 26 February 28, 2024 4.7.2 ENVIRONMENTAL MONITORING
- a. Monitoring of dose on SITE using thermoluminescent (TLD) and/or optically stimulated luminescence (OSL) dosimeters SHALL be performed and documented annually.
- b. Environmental monitoring (e.g., sampling of soil and vegetation) SHALL be performed and documented annually.
4.7.3 STACK MONITOR OPERABILITY Stack particulate activity monitor CHANNEL CHECK SHALL be performed daily when ventilation is required to be operated, and a CHANNEL CALIBRATION annually.
Basis Maintaining the FACILITY in accordance with Surveillance Requirements 4.7.1 and 4.7.3 ensures that the monitoring systems are periodically tested and checked to maintain the instruments OPERABLE.
Based on experience at this SITE, the monitoring frequency of Surveillance Requirement 4.7.2 is adequate to conform to specification 3.7.2.
4.8 EXPERIMENTS These specifications are not applicable due to the reactor being in a POLSC.
Experiments are not authorized.
5 DESIGN FEATURES 5.1 SITE AND FACILITY DESCRIPTION 5.1.1 FACILITY LOCATION The Nuclear Test Reactor (NTR) FACILITY SHALL be located on the SITE of the Vallecitos Nuclear Center (VNC).
5.1.2 CONTROLLED AREA AND RESTRICTED AREA TERMINOLOGY The controlled area, as defined in 10 CFR Part 20 of the Commissions regulations, is the area within the VNC SITE boundary. The restricted area, as defined in 10 CFR Part 20 of the Commissions Regulations, is the NTR FACILITY.
11 Amendment No. 26 February 28, 2024 5.2 PRIMARY COOLANT SYSTEM 5.2.1 PRIMARY SYSTEM PRESSURE The primary coolant system is maintained at atmospheric pressure by a vent line to the holdup tank and the top of the fuel tank being open to the reactor cell.
5.3 REACTOR CORE AND FUEL 5.3.1 POSSESSION ONLY LICENSE SHUTDOWN CONFIGURATION (POLSC)
SAFETY and CONTROL RODS Remain fully inserted and restrained from any movement by ensuring:
SAFETY ROD drive belts removed.
SAFETY RODS electrically isolated.
Course CONTROL ROD drive chains removed.
CONTROL RODS electrically isolated.
MPS MPS of sizes 0.9, 0.5, and a full width poison sheet are installed and latched in slots #1, 2, and 5, respectively.
MPS covers installed, locked and the keys removed from the FACILITY.
Primary coolant Primary coolant system is OPERABLE.
Fuel loading tank is filled with water.
5.3.2 Core Reel Assembly The fuel assemblies SHALL be positioned in a reel assembly inside the core tank. The core reel assembly SHALL be rotated only during authorized fuel handling activities and by manual operation of a crank inside the NTR reactor cell.
5.3.3 Temperature Coefficient of Reactivity The core is designed to exhibit a negative temperature coefficient of reactivity above 124°F, which is approximately the reactor steady-state operating temperature.
12 Amendment No. 26 February 28, 2024 6 ADMINISTRATIVE CONTROLS 6.1 ORGANIZATION The NTR SHALL be owned and maintained by the LICENSEE with the management and support organization as shown in Figure 6-1.
6.1.1 STRUCTURE Figure 6-1 FACILITY Organization 6.1.2 RESPONSIBILITIES (1) The Level 1 LICENSE Holder SHALL be responsible for the NTR FACILITY LICENSE.
(2) The Level 2 Reactor Administrator SHALL ensure that the NTR is maintained according to the FACILITY LICENSE and applicable regulations, and is responsible for security and safety of the FACILITY.
(3) Radiation Safety Function - Radiation Safety Officer is a SITE-wide compliance role that operates independently of the Reactor Administrator and is responsible for safe radiological practices and procedures on the SITE.
(4) The Level 3 Certified Fuel Handler Supervisor is responsible for fuel handling operations and ensures the fuel handling operations are done safely, that staffing is adequate, and that Certified Fuel Handlers have current documented training and qualifications.
13 Amendment No. 26 February 28, 2024 (5) The Level 4 Certified Fuel Handler performs fuel handling operations under the direction of the Certified Fuel Handler Supervisor.
(6) Responsibilities of one level MAY be assumed by alternates when designated in writing.
6.1.3 STAFFING The reactor SHALL not be operated and fuel movement prior to defueling is not permitted. A Certified Fuel Handler Supervisor SHALL be present in the reactor cell during fuel handling operations.
6.1.4 SELECTION AND TRAINING OF PERSONNEL (1) The Reactor Administrator SHALL meet minimal standards for this position that include a cumulative 5 years of reactor experience, with 2 years in an occupational radiation exposure program, and 2 years of personnel supervisory experience. Variations in these standards SHALL be justified in writing by the LICENSE Holder.
(2) Certified Fuel Handler Supervisors and Certified Fuel Handlers SHALL be trained in accordance with the NRC approved Certified Fuel Handler training program for the NTR.
6.2 REVIEW AND AUDIT 6.2.1 COMPOSITION AND QUALIFICATIONS (1) The Oversight Committee SHALL conduct routine audits and perform periodic reviews of the implementation of these Technical Specifications.
(2) The Oversight Committee SHALL be composed of the Level 2 Reactor Administrator and a member of radiation protection staff along with at least three individuals having expertise in reactor technology or radiation protection. Members SHALL be appointed by the Level 1 LICENSE Holder.
6.2.2 CHARTER AND RULES The Oversight Committee SHALL be conducted under a written charter including provisions for:
(1) A meeting frequency of not less than once per calendar year.
(2) Allowing only one vote for each member or alternate for each issue reviewed.
(3) Quorum rules whereby a quorum is at least one-half of the voting members.
14 Amendment No. 26 February 28, 2024 (4) The use of support organizations.
(5) Maintenance of records; including the dissemination, review, and approval of minutes.
6.2.3 REVIEW FUNCTION Activities requiring review SHALL include the following:
(1) Determinations that proposed changes in equipment, systems, tests, or procedures are allowed without prior NRC approval as determined by 10 CFR 50.59 evaluation.
(2) All new procedures and major revisions of existing procedures having safety significance that are required by the administrative control specifications in Section 6.4.
(3) Proposed changes to the Technical Specifications or the FACILITY LICENSE.
(4) Violations of Technical Specifications, and FACILITY LICENSE requirements.
(5) Audit Reports.
6.2.4 AUDIT FUNCTION Audits SHALL include examination of operations records, logs, and documents as well as discussions with staff and observations as appropriate. Deficiencies SHALL be reported to the Level 1 LICENSE Holder as soon as identified and a written report of the findings of the audit submitted to the Oversight Committee within 3 months after the audit has been completed. The following SHALL be audited:
(1) FACILITY activities for conformance to these Technical Specifications and applicable LICENSE conditions: at least once per calendar year not to exceed 15 months between audits.
(2) Certified fuel handling training program: at least once every other calendar year not to exceed 30 months between audits.
(3) The results of condition reports initiated relative to the NTR: once per calendar year not to exceed 15 months between audits.
(4) NTR emergency response implementing procedures: once every other year not to exceed 30 months between audits.
15 Amendment No. 26 February 28, 2024 6.3 RADIATION SAFETY The Level 2 Reactor Administrator (or the Level 3 Certified Fuel Handler Supervisor in his absence), in coordination with the VNC Radiation Safety Officer (RSO), SHALL be responsible for implementing the NTR radiation safety function. The RSO SHALL report relevant findings to the Level 2 Reactor Administrator but SHALL report organizationally to the Level 1 LICENSE Holder, thereby maintaining independence from the production organization. The radiation safety function is informed by the guidelines of the ANSI/ANS 15.11-2016, Radiation Protection at Research Reactor Facilities.
6.4 PROCEDURES Written procedures SHALL be prepared, reviewed, and authorized prior to initiating any of the activities listed in this section. Because the VNC is a multi-LICENSE FACILITY, procedures implementing elements of SITE-wide programs (i.e., radiation protection, emergency planning, security) are authorized by the Level 1 LICENSE Holder. NTR-specific implementing procedures as components of those larger programs and non-administrative changes to those procedures SHALL be authorized by the Level 2 Reactor Administrator.
Procedures exclusive to the implementation of administrative requirements of the NTR Licensing basis and their revisions SHALL be authorized by the Level 2 Reactor Administrator or his designated alternate(s) according to this section. Several of the activities in Section 6.4.1 MAY be included in a single manual or set of procedures or divided among various manuals or procedures.
6.4.1 WRITTEN PROCEDURES Written procedures SHALL be prepared for the following activities as required:
(1) Fuel Handling - Defueling. These may be maintained inactive as fuel handling will not be performed under the POLSC prior to defueling.
(2) Preventive or corrective maintenance which could have an effect on the safety of the fuel in storage, including the replacement of components.
(3) Surveillance checks, tests, calibrations, and inspections required by the Technical Specifications.
(4) NTR-specific radiation protection program implementing procedures for personnel safety consistent with applicable regulations or guidelines.
Management commitment and programs to maintain exposures and releases as low as reasonably achievable SHALL be a component of the SITE-wide
16 Amendment No. 26 February 28, 2024 radiation protection program.
(5) NTR-specific implementing procedures for the SITE-wide emergency and security plans.
(6) NTR-specific radiation protection program implementing procedures for the use, receipt, and on-SITE transfer of by-product material for such activities performed under the R-33 LICENSE.
6.4.2 ADMINISTRATIVE CHANGES TO PROCEDURES (1) ADMINISTRATIVE CHANGES to procedures required by Section 6.4.1 MAY be made by the Level 3 Certified Fuel Handler Supervisor or Level 2 Reactor Administrator before implementation.
(2) ADMINISTRATIVE CHANGES made by authorization of the Level 3 Certified Fuel Handler Supervisor SHALL be subsequently approved by the Level 2 Reactor Administrator.
6.4.3 TEMPORARY DEVIATIONS Temporary deviations from established procedures MAY be made by a Level 3 Certified Fuel Handler Supervisor in order to deal with special or unusual circumstances. These deviations SHALL be documented and reported to the Level 2 Reactor Administrator by the end of the next working day.
6.5 EXPERIMENTS REVIEW AND APPROVAL Experiments are no longer performed at the NTR.
6.6 REQUIRED ACTIONS Actions in response to safety limit violations are not applicable under the POLSC. Reactor operations are not authorized. Internal reporting requirements are included in Section 6.7, Reports.
17 Amendment No. 26 February 28, 2024 6.7 REPORTS 6.7.1 ROUTINE REPORT A routine report providing the following information SHALL be submitted to the NRC Document Control Desk in accordance with the provisions of 10 CFR 50.59 not to exceed 24 months:
(1) Tabulation of major preventive and corrective maintenance activities having safety significance.
(2) A report in accordance with 10 CFR 50.59(d)(2) containing a brief description of any changes, tests, and experiments, including a summary of the evaluation of each.
(3) Summarized results of environmental surveys performed outside the FACILITY.
(4) A summary of exposures received by FACILITY personnel and visitors where such exposures are greater than 25% of that allowed or recommended.
6.7.2 SPECIAL REPORTS Special reports are used to report unplanned events as well as planned major FACILITY and administrative changes. The following special reports SHALL be forwarded to the NRC addressed in accordance with 10 CFR 50.4:
(1) There SHALL be a report not later than the following working day by telephone and confirmed in writing by facsimile or similar conveyance to the NRC Headquarters Operations Officer, to be followed by a written report within 14 days, that describes the circumstances of any of the following events:
- a. Release of radioactivity from the SITE above allowed limits. Such an occurrence SHALL be immediately reported to the Level 2 Reactor Administrator.
- b. Abnormal and significant degradation in reactor fuel, cladding, or coolant boundary, which could result in exceeding prescribed radiation limits for personnel or the environment.
(2) There SHALL be a written report within 30 days to the NRC for:
- a. Permanent changes in the FACILITY organization involving Level 1 or Level 2 management.
- b. Significant changes to the transient or accident analysis as described in the
18 Amendment No. 26 February 28, 2024 CSA 008N0128.
(3) There SHALL be a notification made to the NRC by telephone not later than 10 days following any combination of failures in equipment or administrative radiological work controls that result in a worker being assigned an unplanned dose equal to or greater than 100 mrem Total Effective Dose Equivalent (TEDE) or a spill of more than 1,000 gallons of contaminated liquid waste on uncovered (bare) soil.
6.8 RECORDS Records MAY be in the form of logs, data sheets, or other suitable forms. The required information MAY be contained in single, or multiple records, or a combination thereof.
6.8.1 RECORDS TO BE RETAINED FOR A PERIOD OF AT LEAST FIVE YEARS OR FOR THE LIFE OF THE COMPONENT, WHICHEVER IS LESS (1) Normal reactor FACILITY operation supporting documents (such as checklists, log sheets, etc.,) SHALL be maintained for a period of at least one year.
(2) Principal maintenance activities.
(3) Reportable occurrences.
(4) Surveillance activities required by the Technical Specifications.
(5) Reactor FACILITY radiation and contamination surveys where required by applicable regulations.
(6) Experiments performed with the reactor.
(7) Fuel inventories, receipts, and shipments.
(8) Approved changes in operating procedures.
(9) Records of meeting and audit reports of the review and audit groups.
6.8.2 RECORDS OF THE REQUALIFICATION PROGRAMS Records of the requalification programs SHALL be maintained in accordance with 10 CFR 55.59(c)(5).
19 Amendment No. 26 February 28, 2024 6.8.3 RECORDS TO BE RETAINED FOR THE LIFETIME OF THE REACTOR FACILITY Note: Applicable annual reports, if they contain all the required information, MAY be used as records in this section.
(1) Gaseous and liquid radioactive effluents released to the environs.
(2) Off-SITE environmental-monitoring surveys required by the Technical Specifications.
(3) Radiation exposure for all personnel monitored.
(4) Drawings of the NTR FACILITY.
(5) Changes to the FACILITY made pursuant to 10 CFR 50.59
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 26 TO RENEWED FACILITY OPERATING LICENSE NO. R-33 GE-HITACHI NUCLEAR ENERGY AMERICAS, LLC NUCLEAR TEST REACTOR DOCKET NO. 50-73
1.0 INTRODUCTION
By letter dated October 6, 2023 (Agencywide Documents Access and Management System Accession No. ML23279A110), as supplemented by letters dated December 15, 2023 (ML23349A047), January 26, 2024 (ML24026A196), February 12, 2024 (ML24043A044), and February 23, 2024 (ML24054A539), GE-Hitachi Nuclear Energy Americas, LLC (GEH, the licensee) submitted a license amendment request (LAR) to Renewed Facility Operating License No. R-33, including Appendix A, Technical Specifications [TSs], for the Vallecitos Nuclear Center (VNC) Nuclear Test Reactor (NTR). In the LAR, GEH stated that it commits to permanently cease operation of the NTR and to permanently disable the NRT on or before December 31, 2023. By letter dated January 26, 2024, the licensee confirmed that the NTR had permanently ceased operation on December 21, 2023. As such, GEH requested that Renewed Facility Operating License No. R-33 be amended to remove the authority to operate the NTR, to authorize possession-only of the reactor and reactor fuel, and to remove any operational requirements from the TSs that are not needed for a possession-only license (POL). Further, GEH stated that it will maintain the NTR in a permanently shut down configuration at the VNC location in accordance with the requirements and limitations set forth in the amended license.
2.0 REGULATORY EVALUATION
The U.S. Nuclear Regulatory Commission (NRC, the Commission) staff evaluated the LAR based on the following regulations and guidance:
Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities, Section 50.2, Definitions, which defines a certified fuel handler (CFH) as a non-licensed operator who has qualified in accordance with a fuel handler training program approved by the Commission.
Section 50.36, Technical specifications, of 10 CFR, which requires TSs to be included in utilization facility licenses.
Section 50.54, Conditions of licenses, of 10 CFR, which provides conditions for operating licenses for nuclear reactors.
Section 50.90, Application for amendment of license, construction permit, or early site permit, of 10 CFR, which requires a licensee that desires to amend a license to fully describe the changes desired.
Section 50.120, Training and qualification of nuclear power plant personnel, of 10 CFR, which requires, in part, that the training program for a non-licensed operator must be derived from a systems approach to training, as defined in 10 CFR 55.4, Definitions.
Part 20, Standards for Protection against Radiation, Section 20.1101, Radiation protection programs, of 10 CFR which, in part, requires the licensee to develop, document, and implement a radiation protection program.
Part 55, Operators Licenses, of 10 CFR, which provides the requirements for operators licenses, including 10 CFR 55.4, which defines, among other things, systems approach to training.
NUREG-1537, Part 1, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors: Format and Content, chapter 14, Technical Specifications, appendix 14.1, Format and Content of Technical Specifications for Non-Power Reactors, and chapter 17, Decommissioning and Possession-Only License Amendments, section 17.2, Possession-Only License Amendment (ML042430055),
which provide guidance to research reactor licensees preparing applications for possession-only license amendments.
NUREG-1537, Part 2, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors: Standard Review Plan and Acceptance Criteria, chapter 14, Technical Specifications, and chapter 17, Decommissioning and Possession-Only License Amendments, section 17.2, Possession-Only License Amendment (ML042430048), which provide guidance to the NRC staff for performing reviews of possession-only license amendment requests for research reactors.
American National Standards Institute/American Nuclear Society (ANSI/ANS)-15.1-2007 (R2013), The Development of Technical Specifications for Research Reactors, which provides guidance used by the NRC staff, including the parameters and operating characteristics of a research reactor that should be included in the TSs. Although NUREG-1537, issued in 1996, endorsed ANSI/ANS-15.1-1990, the sections of ANSI/ANS-15.1-2007 (R2013) involved in this review were not changed from ANSI/ANS-15.1-1990.
ANSI/ANS-15.4-2016, Selection and Training of Personnel for Research Reactors, which provides guidance for training research reactor personnel.
ANSI/ANS-15.11-2009, Radiation Protection at Research Reactor Facilities, which provides guidance on the responsibility, resources, and authority to implement the radiation protection program at a research reactor.
3.0 TECHNICAL EVALUATION
3.1 Background
The NTR is a heterogeneous, highly enriched uranium (HEU) fueled, graphite-moderated and reflected, light-water-cooled, thermal reactor licensed to operate at power levels not in excess of 100 kilowatts-thermal (kWt). The fuel consists of HEU-aluminum alloy disks, clad with aluminum. The core is cooled either by natural or forced circulation of deionized light-water circulated in a primary system constructed primarily of aluminum. The reactor operates at low temperature and heat flux. Reactivity is controlled by up to six manually positioned cadmium sheets (MPS), four boron-carbide-filled safety rods (spring-actuated for reactor scram), and three electric motor-driven boron-carbide-filled control rods.
The NTR is located at the VNC, which is largely undeveloped grasslands within the Livermore, California area. The VNC is situated on the north side of Vallecitos Valley in Southern Alameda County within 5 miles of Livermore and Pleasanton and approximately 35 miles east-southeast of San Francisco and 20 miles north of San Jose.
The Atomic Energy Commission issued General Electric (GE) a construction permit for the NTR on October 24, 1957, and an initial facility operating license on October 31, 1957. Renewals to Facility Operating License No. R-33 were issued by License Amendment (LA) No. 18 on December 28, 1984 (ML20111A480 and ML20111A495), and by LA No. 21 on April 20, 2001 (ML003775776). The license was again renewed on June 29, 2023 (ML23128A355).
The NTR was designed and constructed by GE as an experimental physics tool to advance its nuclear energy programs. GEH operated the NTR for neutron radiography of radioactive and nonradioactive objects, small sample irradiation and activation, sensitive reactivity measurements, personnel training, and calibrations and other tests utilizing its neutron flux.
As part of its LAR, the licensee provided enclosure 2, GEH Possession Only License Amendment Safety Analysis of the NTR (Part 1), which provides a safety analysis in support of the proposed POL, and enclosure 3, GEH Possession Only License Amendment Safety Analysis of the NTR (Part 2), which details the criticality safety analysis (CSA) used in support of the proposed POL (the licensee provided a redacted version of this enclosure in its supplemental letter dated December 15, 2023).
The POL LAR safety analysis references the General Electric Nuclear Test Reactor Safety Analysis Report, dated March 24, 2023 (ML23086C028), previously provided to the NRC for use in the most recent license renewal review for the NTR. Any reference to the safety analysis report (SAR) in this safety evaluation (SE) is to the SAR dated March 24, 2023.
The NRC staff notes that by letter dated September 1, 2023 (ML23244A247), GEH requested, in part, approval from the NRC for a direct transfer of control of GEHs licensed interests in Renewed Facility Operating License No. R-33 for the NTR to NorthStar Vallecitos, LLC. This letter indicated that once the transfer is approved by the NRC staff, the NTR has permanently ceased operating and its license has been converted to a POL, and the transfer transaction is consummated, NorthStar Vallecitos, LLC would be the license-holder authorized to conduct licensed activities and to possess the nuclear material at the VNC. NorthStar Vallecitos, LLC plans to complete radiological decommissioning and decontamination of all portions of the VNC, including the NTR, by the end of 2030, with the exception of the Hillside Storage Facility, the specially designed facility licensed to store special nuclear material (SNM) at the site, which will be decommissioned as promptly and as reasonably practicable after removal by the U.S.
Department of Energy of the spent nuclear fuel and high-level radioactive waste located at the site.
NRC Regulatory Audit As part of its review of the LAR, the NRC staff conducted a regulatory audit via teleconference with GEH staff. A regulatory audit plan was provided to GEH by letter dated October 25, 2023 (ML23298A154). During the regulatory audit, the NRC staff issued a request for additional information (RAI), dated November 1, 2023 (ML23305A051). The licensee responded to the RAI by letter dated December 15, 2023. The NRC staff summarized the results of the audit by letter dated February 1, 2024 (ML24031A123).
By letter dated January 26, 2024, the licensee provided an updated version of the proposed TSs. Attachment 1 to that letter, titled, Proposed R-33 NTR Technical Specifications, Revision 8, is referred to hereafter as the proposed TSs, attachment 2 to that letter, titled, Roadmap with the basis and justification for each Technical Specification change, is referred to hereafter as the Roadmap, and attachment 3 to that letter, titled, NTR Certified Fuel Handler Training and Requalification Program (CFHTRP), is referred to hereafter as the CFHTP.
On February 8, 2024, during a telephone conversation following the audit, the licensee indicated that item no. 3 in the cover letter to the LAR, which states that [t]he control room console key will be placed in the off position, removed, and be locked in a secured location away from the facility, was not necessary as it would not be possible to restart the reactor from the console following the implementation of the POL TSs. Thereafter, by letter dated February 12, 2024, the licensee provided its justification for no longer considering the requirements in item no. 3 as necessary. The NRC staffs review finds that the requirements in proposed TS 3.2.1, TS 3.2.2, TS 4.2.1, TS 4.2.2, and TS 5.3.1 will ensure that reactor operation is not possible and that, therefore, the requirements proposed in LAR item no. 3 are not necessary.
By letter dated February 23, 2024 (ML24054A539), the licensee withdrew its proposed deletion of License Condition 1.C, as stated in the LAR dated October 6, 2023, and provided a justification for proposed TS 6.8.3(5). In enclosure 1 to that letter, the licensee provided a revision of the CFHTP, titled, NTR Certified Fuel Handler Training Program (CFHTP),
Revision 1, which replaced the initial CFHTP provided by letter dated January 26, 2024, in its entirety.
Radiological Emergency Plan In the LAR, enclosure 2, the licensee stated that there were no plans to revise the VNC Radiological Emergency Plan (REP) as a result of the proposed POL. The NRC staff notes that it recently reviewed and approved the VNC REP in General Electric-Hitachi Nuclear Energy Americas, LLC - Issuance of Amendment No. 24 to Facility License No. R-33 for Vallecitos Nuclear Center Radiological Emergency Plan, dated February 12, 2021 (ML20125A077).
Further, the licensee recently revised the VNC REP in accordance with the provisions of 10 CFR 50.54(q), as stated in its letter dated October 27, 2022 (ML22300A210).
Physical Security Plan In the LAR, enclosure 2, the licensee stated that a revision is planned to the physical security plan (PSP) in support of the LAR. However, after discussion with the NRC staff, the licensee clarified the statement, as provided in the Roadmap, section A, Storage of Fuel, item no. 6, which indicated that no changes to the PSP were needed prior to the issuance of the LAR, and the changes requested in the LAR are future changes that will be made in accordance with the requirements of 10 CFR 50.54(p) and will not need prior NRC review or approval.
3.2 Proposed Changes to Renewed Facility Operating License No. R-33 for the NTR In the LAR, the licensee proposed changes to Renewed Facility Operating License No. R-33 for the NTR to reflect the permanently shutdown and, therefore, possession-only status of the facility. In its response to RAI No. 1, the licensee acknowledged acceptance of the POL license conditions (LCs) proposed by the NRC staff. In addition, by letter dated February 23, 2024, the licensee requested to withdraw its deletion of LC 1.C. The current LCs and the final proposed LCs are provided below along with the NRC staffs evaluation of the changes.
Current LC 2.B.(1) states:
(1) Pursuant to Section 104c of the Act and 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, to possess, use, and operate the reactor as a utilization facility at the designated location in Alameda County, California, in accordance with the procedures and limitations described in the application and set forth in this license.
Proposed LC 2.B.(1) states:
(1) Pursuant to Section 104c of the Act and 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, to possess the reactor at the designated location in Alameda County, California, in accordance with the procedures and limitations described in the application and set forth in this license.
In the LAR, the licensee indicated that it would no longer operate the facility at any power level because it planned to permanently cease operation on or before December 31, 2023, and did, in fact, permanently cease operation on December 21, 2023.
The NRC staffs review finds the proposed change to LC 2.B.(1) consistent with the possession-only status of the NTR as it removes the authority to operate the reactor. The NRC staff also finds the proposed change consistent with the guidance for possession-only LCs provided in NUREG-1537, Part 1, section 17.2.1.1. Therefore, the NRC staff concludes that proposed LC 2.B.(1) is acceptable.
Current LC 2.B.(2) states:
(2) Pursuant to the Act and 10 CFR Part 70, to receive, possess, and use in connection with the operation of the reactor:
- a. but not separate, up to 4 kilograms of contained uranium-235 of an enrichment of 20 percent or greater in the isotope uranium-235, in the form of in-core reactor fuel;
- b. but not separate, up to 350 grams of contained uranium-235, that is not to be used as in-core fuel. This material can be used in the reactor cell, south cell, north room, and control room but not in the experimental facilities of the NTR.
- c. but not separate, such special nuclear material as may be produced by the operation of the reactor.
Proposed LC 2.B.(2) states:
(2) Pursuant to the Act and 10 CFR Part 70, to possess, but not separate:
- a. up to 4 kilograms of contained uranium-235 of an enrichment of 20 percent or greater in the isotope uranium-235, in the form of in-core reactor fuel; and
- b. such special nuclear material as may have been produced by the operation of the reactor.
In the LAR, the licensee stated that all SNM and non-reactor-related byproduct material has been removed from the NTR facility except for the fuel in storage in the reactor.
The NRC staffs review finds the proposed change to LC 2.B.(2) consistent with the licensees statement that SNM will be remaining at the facility during the possession-only status of the NTR license. The NRC staff also finds the proposed change consistent with the guidance for possession-only LCs provided in NUREG-1537, Part 1, section 17.2.1.1. Therefore, the NRC staff concludes that proposed LC 2.B.(2) is acceptable.
Current LC 2.B.(3) states:
(3) Pursuant to the Act and 10 CFR Part 30, to receive, possess, and use in connection with the operation of the facility:
- a. up to 2,000 curies of either activated solids as contained in such items as encapsulating materials, structural material and irradiated components;
- b. up to 0.2-curie radium-beryllium sealed startup source, and up to 0.03 curies of byproduct materials, in the form of sealed sources, for instrument calibration and source checks; and
- c. but not to separate (except for byproduct material produced as allowed for experiments), such byproduct material as may be produced by the operation of the reactor.
Proposed LC 2.B.(3) states:
(3) Pursuant to the Act and 10 CFR Part 30, to possess:
- a. up to 0.2-curie radium-beryllium sealed startup source, and up to 0.03 curies of byproduct materials, in the form of sealed sources, for instrument calibration and source checks; and
- b. but not to separate such byproduct material as may have been produced by the operation of the reactor.
In the LAR, the licensee stated that all SNM and non-reactor-related byproduct material has been removed from the NTR facility except for the fuel in storage in the reactor.
The NRC staffs review finds the proposed change to LC 2.B.(3) consistent with the licensees statement that byproduct material will be remaining at the facility during the possession-only status of the NTR license. The NRC staff also finds the proposed change consistent with the guidance for possession-only LCs provided in NUREG-1537, Part 1, section 17.2.1.1.
Therefore, the NRC staff concludes that proposed LC 2.B.(3) is acceptable.
Current LC 2.C.(1) states:
(1) Maximum Power Level The licensee is authorized to operate the facility at steady-state power levels not in excess of 100 kilowatts (thermal) in accordance with the limitations in the Technical Specifications.
Proposed LC 2.C.(1) states:
(1) Maximum Power Level The licensee is not authorized to operate the reactor at any power level.
In the LAR, the licensee indicated that it would no longer operate the facility at any power level because it planned to permanently cease operation on or before December 31, 2023, and did, in fact, permanently cease operation on December 21, 2023.
The NRC staffs review finds the proposed change to LC 2.C.(1) consistent with the possession-only status of the NTR as it removes the authority to operate the reactor. The NRC staff also finds the proposed change consistent with the guidance for possession-only LCs provided in NUREG-1537, Part 1, section 17.2.1.1. Therefore, the NRC staff concludes that proposed LC 2.C.(1) is acceptable.
Current LC 2.C.(2) states:
(2) Technical Specifications The Technical Specifications contained in Appendix A are hereby incorporated in their entirety in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
Proposed LC 2.C.(2) states:
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised by Amendment No. 26, are hereby incorporated in their entirety in the license. The licensee shall maintain the facility in accordance with the Technical Specifications.
In the LAR, the licensee proposed to add to this LC the phrase as revised by Amendment No. 26. The NRC staff finds that the proposed change is acceptable as Amendment No. 26 will be assigned, in sequential order, to this amendment as part of the NRCs license amendment process.
In its response to RAI No. 1, the licensee requested that operate be changed to maintain to reflect the possession-only status of the NTR. The NRC staffs review finds the proposed change to LC 2.C.(2) consistent with the possession-only status of the NTR as it removes the authority to operate the reactor. The NRC staff also finds the proposed change consistent with the guidance for possession-only LCs provided in NUREG-1537, Part 1, section 17.2.1.1.
Therefore, the NRC staff concludes that proposed LC 2.C.(2) is acceptable.
Current LC 2.D states:
This license is effective as of the date of its issuance and shall expire at midnight, 20 years from its date of issuance.
Proposed LC 2.D states:
The license is effective as of the date of its issuance and until the Commission notifies the licensee in writing that the license is terminated.
The proposed change would remove the license expiration date and replace it with a statement summarizing the relevant requirement of 10 CFR 50.51(b). Specifically, that regulation states, in part, that [e]ach license for a facility that has permanently ceased operations, continues in effect beyond the expiration date to authorize ownership and possession of the production or utilization facility, until the Commission notifies the licensee in writing that the license is terminated. Because proposed LC 2.D is consistent with the NRCs regulations and the status of the NTR as permanently shut down, the NRC staff concludes that it is acceptable.
3.3 Proposed Changes to the Technical Specifications for the NTR In the LAR, the licensee proposed changes to the TSs for the NTR to reflect the possession-only status of the facility. For clarity, since many of the proposed changes involve deleted text, this SE does not specifically identify the text that the licensee proposed to delete, but rather describes those deletions, as necessary, in the discussion regarding the proposed TS changes.
The remaining changes proposed for each TS are indicated by bold font.
3.3.1 TS 1, Introduction Current TS 1, Introduction, states, in part:
These Technical Specifications provide limits within which operation of the reactor will assure the health and safety of the public, the environment, and on-SITE personnel. Areas addressed are Definitions, Safety Limits (SL),
Limiting Safety System Settings (LSSS), Limiting Conditions for Operation (LCO), Surveillance Requirements, Design Features and Administrative Controls.
Proposed TS 1, Introduction, states, in part:
These Technical Specifications provide limits that assure reactor-related activity will be controlled in a way that protects the health and safety of the public, the environment, and on-SITE personnel. Areas addressed are Definitions, Limiting Conditions for POLSC (LCP), Surveillance Requirements, Design Features, and Administrative Controls.
In the Roadmap, section C, Changes to NTR Technical Specifications, item 1, Section 1, Introduction, the licensee states that the verbiage was changed to reflect that the reactor will not be operated but reactor-related activity will be controlled in a way that protects the health and safety of the public, environment, and on-site personnel.
The NRC staffs review finds the licensees proposed changes to TS 1 consistent with the POL status of the facility. The NRC staff also finds the proposed changes consistent with the review guidance in NUREG-1537, Part 2, section 17.2.1.2, Technical Specifications. Therefore, the NRC staff concludes that proposed TS 1 is acceptable.
3.3.2 TS 1.1, Definitions Current TS Definition, Control Rod(s), states:
A non-scrammable device having an electric motor drive. The rod contains boron-carbide material used to establish neutron flux changes and to compensate for routine reactivity losses (Refer to Design Feature 5.3.1.).
Proposed TS Definition, Control Rods(s), states:
A non-scrammable device having an electric motor drive and containing boron-carbide material. These rods have been disabled and remain fully inserted and restrained from any movement in the core per the POSSESSION ONLY LICENSE SHUTDOWN CONFIGURATION.
In the Roadmap, section C, item 2, Control Rods, the licensee states that the function of the control rods has changed to reflect their function under the possession only license shutdown configuration (POLSC) to be fully and permanently inserted into the core and restrained from movement to maintain the reactor subcritical according to the criticality safety analysis.
The NRC staffs review finds the licensees proposed changes to the TS definition for Control Rod(s) consistent with the proposed POL status of the facility and with the requirements of the POLSC stated in proposed TS 5.3.1, Possession Only License Shutdown Configuration (POLSC). The NRC staff evaluated proposed TS. 5.3.1 and finds the proposed TS acceptable (see SE section 3.3.6, TS 5, Design Features). The NRC staff also finds the proposed changes consistent with the review guidance in NUREG-1537, Part 2, section 17.2.1.2.
Therefore, the NRC staff concludes that the proposed TS definition for Control Rod(s) is acceptable.
Current TS Definition, Core Configuration, states:
The fixed assembly that includes 16 fuel assemblies each containing 40 fuel discs. The assemblies are contained within and evenly distributed around the annular core tank (Refer to Design Feature TS 5.3.1). Positioned around the outer edge of the core tank are four SAFETY RODS, three CONTROL RODS, and installed MANUAL POISON SHEETS.
Proposed TS Definition, Core Configuration, states:
See POSSESSION ONLY LICENSE SHUTDOWN CONFIGURATION.
In the Roadmap, section C, item 2, Core Configuration, the licensee states that the definition has been changed to reflect that the sole core configuration under the POL is according to the POLSC.
The NRC staffs review finds the licensees proposed changes to the TS definition for Core Configuration consistent with the POLSC as described in proposed TS 5.3.1. The NRC staff evaluated proposed TS.5.3.1 and finds the proposed TS acceptable (see SE section 3.3.6). The NRC staff also finds the proposed changes consistent with the review guidance in NUREG 1537, Part 2, section 17.2.1.2. Therefore, the NRC staff concludes that the proposed TS definition for Core Configuration is acceptable.
Current TS Definition, Manual Poison Sheet(s) (MPS), states:
Manually positioned devices containing cadmium material used to compensate for fuel burnout and limit the amount of POTENTIAL EXCESS REACTIVITY available to the operator (Refer to Design Feature 5.3.1).
Proposed TS Definition, Manual Poison Sheet(s) (MPS), states:
Manually positioned devices containing cadmium material used to maintain adequate negative reactivity inventory in the reactor to prevent attainment of criticality.
In the Roadmap, section C, item 2, Manual Poison Sheets (MPS), the licensee states that the function of the MPS has changed to that of ensuring that the reactor remains subcritical.
The NRC staffs review finds the licensees proposed changes to the TS definition for MPS consistent with the POLSC as described in proposed TS 5.3.1. The NRC staff evaluated proposed TS 5.3.1 and finds the proposed TS acceptable (see SE section 3.3.6). The NRC staff also finds the proposed changes consistent with the review guidance in NUREG 1537, Part 2, section 17.2.1.2. Therefore, the NRC staff concludes that the proposed TS definition for MPS is acceptable.
Proposed new TS Definition, Possession Only License Shutdown Configuration (POLSC),
states:
That plant configuration that ensures the reactor will remain subcritical in the credible limiting accident analysis by restraining CONTROL RODS, SAFETY RODS, and MPS in the positions assumed in the criticality safety analysis. Details of the NTR POLSC are included in Section 5.
In the Roadmap, section C, item 2, Possession Only License Shutdown Configuration (POLSC), the licensee states that the proposed new definition of POLSC was added to describe the facility modifications needed to maintain the reactor in the shutdown configuration, including the proposed TSs in TS 5.
The NRC staffs review finds the licensees proposed new TS definition for POLSC consistent with the proposed TS 5.3.1. The NRC staff evaluated proposed TS 5.3.1 and finds the proposed TS acceptable (see SE section 3.3.6). The NRC staff also finds the proposed new TS definition for POLSC consistent with the review guidance in NUREG 1537, Part 2, section 17.2.1.2.
Therefore, the NRC staff concludes that the proposed new TS definition for POLSC is acceptable.
Current TS Definition, Reactor Shutdown, states:
The reactor is shutdown if it is subcritical by at least one dollar in the REFERENCE CORE CONDITION with the REACTIVITY WORTH of all installed EXPERIMENTs included.
Proposed TS Definition, Reactor Shutdown Configuration (retitled from Reactor Shutdown),
states:
All SAFETY RODS, CONTROL RODS, and in-service MANUAL POISON SHEETS SHALL remain in their respective fully inserted positions. See POLSC.
In the Roadmap, section C, item 2, Reactor Shutdown Configuration, the licensee states that the definition was proposed to be modified to describe the core configuration (rods and MPS fully inserted) in support of the criticality safety analysis to maintain the reactor subcritical.
The NRC staffs review finds the licensees proposed changes to the TS definition for Reactor Shutdown/Reactor Shutdown Configuration consistent with the criticality safety analysis, reviewed and found acceptable by the NRC staff in SE section 3.5, Criticality Safety Analysis.
The NRC staff also finds the proposed changes consistent with the review guidance in NUREG 1537, Part 2, section 17.2.1.2. Therefore, the NRC staff concludes that the proposed TS definition for Reactor Shutdown Configuration is acceptable.
Current TS Definition, Safety Rod(s), states:
Spring-actuated scrammable devices containing boron-carbide material used to perform the safety function of ensuring the reactor can be placed in REACTOR SHUTDOWN from any OPERATING condition. (Refer to Design Feature 5.3.1).
Proposed TS Definition, Safety Rod(s), states:
Previously scrammable devices containing boron-carbide material. These devices have been disabled, are fully inserted, and restrained from any movement, and remain in the core per the POLSC.
In the Roadmap, section C, item 2, Safety Rods, the licensee states that the function of the safety rods has been changed to reflect their function under the POLSC to be fully and permanently inserted into the core and restrained from movement to maintain the reactor subcritical according to the criticality safety analysis.
The NRC staffs review finds the licensees proposed changes to the TS definition for Safety Rod(s) consistent with the possession-only status and the POLSC as described in proposed TS 5.3.1. The NRC staff evaluated proposed TS 5.3.1 and finds the proposed TS acceptable (see SE section 3.3.6). The NRC staff also finds the proposed changes to the TS definition for Safety Rod(s) consistent with the criticality safety analysis, reviewed and found acceptable by the NRC staff in SE section 3.5. Further, the NRC staff finds the proposed changes consistent with the review guidance in NUREG-1537, Part 2, section 17.2.1.2. Therefore, the NRC staff concludes that the proposed TS definition for of Safety Rod(s) is acceptable.
Additionally, the licensee proposed to delete the current TS definitions listed below, since the NTR with the approval of the LAR will no longer be authorized to operate or to conduct experiments:
EXPERIMENT(S)
EXPLOSIVE MATERIAL FLAMMABLE LICENSED REACTOR OPERATOR(S)/REACTOR OPERATOR(S)/SENIOR REACTOR OPERATOR(S)
MEASURED VALUE POTENTIAL EXCESS REACTIVITY PROTECTIVE ACTION(S)
REACTIVITY WORTH (EXPERIMENT)
REACTOR OPERATING or REACTOR OPERATION(S)
REACTOR THERMAL POWER REACTOR SAFETY SYSTEM(S)
REACTOR SECURED REFERENCE CORE CONDITION SCRAM TIME SHUTDOWN MARGIN SURVEILLANCE INTERVALS:
Quinquennial - interval not to exceed 70 months.
Monthly - interval not to exceed 6 weeks.
Weekly - interval not to exceed 10 days.
Prior to SU - Prior to the first reactor start-up of the day.
TRUE VALUE UNSAFE CONDITION UNSCHEDULED SHUTDOWN(S)
The NRC staffs review finds the licensee's proposed deletions, listed above, to the current TS definitions consistent with the possession-only status of the facility in that these definitions are not needed for a permanently shut down facility. Further, the NRC staff finds the proposed deletions consistent with the review guidance provided in NUREG-1537, Part 2, section 17.2.1.2. Therefore, the NRC staff concludes that the proposed TS definitions deletions are acceptable.
3.3.3 TS 2, Safety Limits and Limiting Safety System Settings Current TS 2.1, Safety Limits, states:
REACTOR THERMAL POWER The TRUE VALUE of the REACTOR THERMAL POWER SHALL not exceed 190 kW.
Proposed TS 2.1, Safety Limits, states:
These specifications are not applicable due to the reactor being in a POLSC. Reactor operations are not authorized.
Current TS 2.2, Limiting Safety System Settings, states:
Linear Power - MEASURED VALUE The linear neutron power monitor CHANNEL set point SHALL not exceed the MEASURED VALUE of 125 kW.
Proposed TS 2.2, Limiting Safety System Settings, states:
These specifications are not applicable due to the reactor being in a POLSC. Reactor operations are not authorized.
In the Roadmap, section C, item 3, Section 2.1, Safety Limits, and item 4, Section 2.2, Limiting Safety System Settings, the licensee states that current TS 2.1 and TS 2.2 are not consistent with a reactor that has permanently shut down, and proposes revised TS 2.1 and TS 2.2 to reflect the possession-only status.
The NRC staffs review finds the licensees proposed changes to TS 2.1 and TS 2.2 consistent with the possession-only status of the facility. Further, the NRC staff finds the proposed changes consistent with the review guidance in NUREG-1537, Part 2, section 17.2.1.2.
Therefore, the NRC staff concludes that proposed TS 2.1 and TS 2.2 are acceptable.
3.3.4 TS 3, Limiting Conditions for Operation (LCO)
Current TS 3.1, Reactor Core Parameters, states:
3.1.1 POTENTIAL EXCESS REACTIVITY POTENTIAL EXCESS REACTIVITY SHALL be $0.76. If it is determined to be > $0.76, the reactor SHALL be placed in REACTOR SHUTDOWN immediately.
3.1.2 SUBCRITICAL ROD POSITION The reactor SHALL be subcritical whenever the four SAFETY RODS are withdrawn from the core and the three CONTROL RODS are fully inserted. Place reactor in REACTOR SHUTDOWN if this condition is not met.
3.1.3 MINIMUM SHUTDOWN MARGIN The minimum SHUTDOWN MARGIN with the maximum worth SAFETY ROD stuck out SHALL be $1.0.
Proposed TS 3.1, Reactor Core Parameters, states:
3.1 REACTOR CORE PARAMETERS Reactor operations are not authorized. Fuel handling in support of defueling is the only activity allowed.
In the Roadmap, section C, item 5, Section 3, Limiting Conditions for Operation (LCO), the licensee proposed to retitle TS 3 from Limiting Conditions for Operation (LCO) to Limiting Conditions for POLSC (LCP) to reflect that the reactor has permanently ceased operation.
The NRC staffs review finds the licensees proposed change to the title consistent with the possession-only status of the NTR and, therefore, acceptable.
In the Roadmap, section C, item 6, Section 3.1, Reactor Core Parameters, the licensee states that current TS 3.1 was not consistent with a reactor that has permanently shut down, and proposed to revise TS 3.1 to reflect the possession-only status. The licensee also states that fuel handling in support of defueling is the only fuel handling activity allowed during the possession-only status.
The NRC staffs review finds the licensees proposed changes to TS 3.1 consistent with the possession-only status of the facility. The NRC staff also finds the proposed changes consistent with the review guidance in NUREG-1537, Part 2, section 17.2.1.2. Therefore, the NRC staff concludes that proposed TS 3.1 is acceptable.
Current TS 3.2, Reactor Control and Safety System, states:
3.2.1 RODS OPERABLE REACTOR OPERATION SHALL be permitted only when all four SAFETY RODS and all three CONTROL RODS are OPERABLE. The reactor SHALL be placed in REACTOR SHUTDOWN immediately if it is known that a SAFETY ROD or CONTROL ROD is NOT OPERABLE.
3.2.2 SAFETY ROD WITHDRAWAL No more than one SAFETY ROD SHALL be simultaneously moved in an outward direction.
3.2.3 SAFETY ROD WITHDRAWAL RATE The rate of withdrawal of each SAFETY ROD during REACTOR OPERATION SHALL be less than 1 1/4 inches per second.
3.2.4 CONTROL ROD WITHDRAWAL RATE The rate of withdrawal of CONTROL RODS during REACTOR OPERATION SHALL be less than 1/6 inch per second. The rods can be inserted or withdrawn singly or multiple rods simultaneously.
3.2.5 SCRAM TIME The average SCRAM TIME of the four SAFETY RODS SHALL not exceed 300 msec.
3.2.6 REACTOR SAFETY SYSTEM AND SAFETY-RELATED ITEMS REACTOR OPERATION SHALL be permitted only when the REACTOR SAFETY SYSTEM is OPERABLE in accordance with Table 3-1 and Table 3-2.
Table 3-1 specifies automatic trip set points, scram system components, and minimum number of CHANNELS necessary to ensure PROTECTIVE ACTIONS can be taken to place the reactor in REACTOR SHUTDOWN. The Trip Points in Table 3-1 reflect the minimum values necessary to avoid approaching the LCOs in Sections 3.1 and 3.2 of these Technical Specifications.
Table 3-2 specifies alarm set points and rod interlock features that prompt operator actions that ensure the FACILITY is maintained within normal OPERATING parameters.
Table 3-1 REACTOR SAFETY SYSTEM - SCRAM Item No.
System Condition Trip Point Function Min.
Number of Channels
- 1.
Linear Power High reactor power 125 kW Scram (2-out-of-3 or 1-out-of-2) 2 Loss of positive high voltage to ion chambers (if used)
No less than 90% of OPERATING voltage Scram (2-out-of-3 or 1-out-of-2)
- 2.
Log N Fast reactor period No less than +5 sec Scram 1
Amplifier Mode switch not in operate N/A Scram Loss of positive high voltage to ion chambers (if used)
No less than 90% of OPERATING voltage Scram
- 3.
Primary Coolant Temperature (Fenwall)
High core outlet temperature 222 °F Scram 1
- 4.
Primary Coolant Flow Low Flow No less than 15 gpm when reactor power >
0.1 kW Scram 1
- 5.
Manual Console button depressed N/A Scram 1
- 6.
Electrical Power Reactor console key in off position (loss of AC power to console)
N/A Scram 1
Table 3-2 REACTOR SAFETY-RELATED ITEMS Item No.
System Condition Set Point Function
- 1.
Reactor Cell Pressure Low Differential pressure
> 0.5 in. water P
Visible and audible alarm; audible alarm MAY be bypassed after recognition.
2 Fuel Loading Tank Water Level Low Level
< 3 ft. below the overflow Visible and audible alarm; audible alarm MAY be bypassed after recognition.
- 3.
Primary Coolant Temperature High core outlet temperature
<200°F Visible and audible alarm; audible alarm MAY be bypassed after recognition.
- 4.
Primary Coolant Temperatures Core Delta temperature N/A Provide information for the heat balance determination
- 5.
Stack Radioactivity High Level Complies with TS 3.7.2.1 Visible and audible alarm; audible alarm MAY be bypassed after recognition.
- 6.
Linear Power Low Power indication 2% on any scale SAFETY RODS or CONTROL RODS cannot be withdrawn 2-out-of-3 or 1-out-of-2).
- 7.
CONTROL ROD or SAFETY ROD Rods not in N/A SAFETY ROD magnets cannot be reenergized
- 8.
SAFETY ROD Rods not out N/A CONTROL RODS cannot be withdrawn; SAFETY RODS SHALL be withdrawn in sequence; MAY be bypassed to allow withdrawal of one CONTROL ROD, or one SAFETY ROD (drive) out of sequence for purposes of inspection, maintenance, and testing Proposed TS 3.2, Reactor Control and Safety System, states:
3.2.1 RODS INOPERABLE All SAFETY RODS and CONTROL RODS SHALL be maintained fully inserted and restrained from any movement per the applicable conditions defined by the POLSC in TS 5.3.1.
3.2.2 MANUAL POISON SHEETS SECURED MPS slots SHALL be maintained per the applicable conditions defined by the POLSC in TS 5.3.1.
In the Roadmap, section C, item 7, Section 3.2, Reactor Control and Safety Systems, the licensee states that the title for TS 3.2.1 was proposed to be changed from Rods Operable to Rods Inoperable to ensure that the reactor shutdown configuration was consistent with the requirements of proposed TS 5.3.1. Also, the licensee states that current TS 3.2.2 was proposed to be revised to include the MPS to reflect the change in the purpose of the MPS during the possession-only status to ensure that the reactor remains shutdown. Further, the licensee states that current TSs 3.2.3 through 3.2.6 and TS tables 3-1 and 3-2 were proposed to be deleted as they were not consistent with the possession-only status of the NTR.
The NRC staffs review finds the licensees proposed changes to TSs 3.2.1 and 3.2.2 consistent with proposed TS 5.3.1, and that the proposed changes will help ensure that the reactor remains shutdown during the possession-only status. The NRC staff evaluated proposed TS 5.3.1 and finds the proposed TS acceptable (see SE section 3.3.6). The NRC staff also finds deleting TSs 3.2.3 through 3.2.6 and TS tables 3-1 and table 3-2 consistent with the possession-only status of the NTR. In addition, the NRC staff finds the proposed changes consistent with the guidance in NUREG-1537, Part 2, section 17.2.1.2. Therefore, the NRC staff concludes that proposed TS 3.2.1 and TS 3.2.2 are acceptable.
Current TS 3.3, Reactor Coolant System, states:
3.3.1 FORCED FLOW COOLING For REACTOR OPERATION above 0.1 kW, the reactor SHALL be cooled by light water forced coolant flow in REACTOR OPERATING mode.
3.3.2 CORE TANK FULL REACTOR OPERATION SHALL not be permitted unless the fuel loading tank is filled with water which ensures that the core tank is full. If during operation of the reactor it is determined that the fuel loading tank is not filled with water, the reactor SHALL be placed in REACTOR SHUTDOWN immediately.
3.3.3 PRIMARY COOLANT CONDUCTIVITY The specific conductivity of the primary coolant water SHALL be maintained less than 5 µS/cm when averaged over a one-month period.
Proposed TS 3.3, Primary Coolant System, states:
3.3.1 FORCED FLOW COOLING Forced primary coolant flow SHALL be OPERABLE. If this condition is not met, corrective action SHALL be taken to restore operability within 90 days.
3.3.2 FUEL LOADING TANK FULL The fuel loading tank SHALL be maintained above the low level alarm.
3.3.3 FUEL LOADING TANK LEVEL ALARM The fuel loading tank low level alarm SHALL be maintained at 3 feet below the overflow or higher. A local visible and remote alarm SHALL be available.
Corrective action SHALL be taken if this condition is not met.
3.3.4 PRIMARY COOLANT CONDUCTIVITY The specific conductivity of the primary coolant water SHALL be maintained less than 5 µS/cm when averaged over the four most recent quarterly readings.
Corrective action SHALL be taken if this condition is not met.
In the LAR, enclosure 2, section D, POLSC System Operability, Primary Coolant System, the licensee states that the reactor fuel will remain in the reactor during the POL period. The POLSC includes the need to routinely (but not continuously) operate the primary coolant system to provide chemistry control for the fuel cladding. Flow through the system is necessary for the operation of the conductivity meter and to ensure mixing of chemistry throughout the system.
In the Roadmap, section C, item 8, Section 3.3, Reactor Coolant System, the licensee states that the title for TS 3.3 was proposed to be changed from Reactor Coolant System to Primary Coolant System to better align with the terminology in the SAR. The licensee also states that:
Current TS 3.3.1, Forced Flow Cooling, was proposed to be edited to remove the cooling function as it is not needed for cooling since the reactor is no longer operating.
The licensee also proposed a 90-day period to restore system operability.
Current TS 3.3.2, Core Tank Full, was proposed to be retitled to Fuel Loading Tank Full because the Fuel Loading Tank is at a higher elevation than the Core Tank in the system, so monitoring its water level will ensure that the Core Tank is full of primary coolant water.
Proposed TS 3.3.3, Fuel Loading Tank Level Alarm, was proposed to be added to relocate the requirements from current TS table 3-2, Reactor Safety-Related Items, item no. 2, Fuel Loading Tank Level Alarm, to help ensure that water level is monitored and maintained.
Current TS 3.3.3, Primary Coolant Conductivity, was proposed to be renumbered to TS 3.3.4, and revised to average the conductivity readings over a quarterly periodicity because of the less harsh conditions, relative to reactor operation, in the possession-only status.
The NRC staffs review of the licensees proposed changes to TS 3.3.1 finds that routine operation of the reactor/primary coolant system is appropriate for chemistry control. Further, the 90-day requirement to restore forced flow operability will help to ensure the availability of the reactor/primary coolant system to maintain chemistry control.
The NRC staffs review finds that the licensees proposed changes to TSs 3.3.2 and 3.3.3 will help maintain the fuel loading tank and ensure that the core tank remains full of primary coolant water, thus establishing a means to control the water chemistry in support of preserving the fuel cladding. Further, the full loading tank alarm in proposed TS 3.3.3 remains unchanged from the current TS table 3-2, item no. 2, and is only relocated within the POL TSs for clarity. The NRC staffs review finds that the proposed change to TS 3.3.4, to average the conductivity readings over four (4) quarters, is acceptable for the possession-only status of the NTR, as less activation products will be produced with the reactor permanently shut down.
Based on the above, the NRC staff concludes that proposed TSs 3.3.1 through 3.3.4 are acceptable.
Current TS 3.5, Reactor Cell, Ventilation, and Confinement System, states:
3.5.1 REACTOR CELL NEGATIVE PRESSURE In REACTOR OPERATING mode, reactor power SHALL not be increased above 0.1 kW unless the reactor cell is maintained at a negative pressure of not less than 0.5 in. of water with respect to the control room.
IF during operation of the reactor above 0.1 kW, the negative pressure with respect to the control room is not maintained, then the reactor power SHALL be lowered to less than 0.1 kW immediately.
3.5.2 REACTOR CELL ACTIVITY RELEASE Reactor cell ventilation system SHALL be OPERATING during performance of activities that could release airborne radioactivity into the reactor cell.
Proposed TS 3.5, Reactor Cell, Ventilation, and Confinement System, states:
3.5.1 REACTOR CELL NEGATIVE PRESSURE The ventilation system SHALL be OPERATING during evolutions that could result in an airborne concentration of one DAC or greater in the reactor cell. A reactor cell negative differential pressure of not less than 0.5 in. of water with respect to the control room SHALL be verified prior to commencing the evolution.
In the Roadmap, section C, item 9, Section 3.5, Reactor Cell, Ventilation, and Confinement System, the licensee states that the proposed changes reflect that the system now serves a primary function of minimizing airborne contaminants in the facility to below one derived air concentration (DAC) of airborne radioactivity by providing a filtered pathway to evacuate airborne radioactive contaminants to the outside environment. Further, the criticality safety analysis indicates that the system will ensure that any emissions will not exceed the limits in 10 CFR 20.1301, Dose limits for individual members of the public.
The NRC staffs review finds that the licensees proposed changes to TS 3.5.1 will continue to ensure that the ventilation system remains available during the possession-only status of the NTR to help control airborne radioactivity in the reactor cell and limit releases to the environment to within the NRCs limits. Therefore, the NRC staff concludes that proposed TS 3.5 is acceptable.
Current TS 3.7, Radiation Monitoring Systems and Effluents, states:
3.7.1 MONITORING SYSTEMS DURING REACTOR OPERATIONS Functional area radiation monitors* are required in EXPERIMENTAL FACILITY spaces while EXPERIMENTS are in progress and the control room during REACTOR OPERATIONS.
3.7.2 MONITORING SYSTEMS DURING REACTOR CELL MAINTENANCE A functional area radiation monitor* is required in the reactor cell during maintenance activities.
- A functional area radiation monitor SHALL include:
Instrument readout that is visible in the control room.
a gamma-sensitive instrument.
A local audible alarm.
3.7.3 EFFLUENTS - ENVIRONMENTAL MONITORING The VNC SITE utilizes environmental air sampling stations and TLD badges in locations specified by the VNC Environmental Monitoring Manual.
3.7.4 EFFLUENTS - STACK RELEASE ACTIVITY The stack discharge rates of gaseous and particulate activity SHALL not exceed the limits in Table 3-3, ensuring compliance with the 10 CFR 20.1101(d) limit 10 mrem/year.
Table 3-3 STACK RELEASE ACTION LEVELS Gaseous Activity (Ar-41)
Particulate Activity (Beta)
Weekly release 9 Ci/wk 1.7+03 µCi/wk Alarm setpoint 9.5E-05 µCi/cc 1.9E-08 µCi/cc
- 1. If the alarm setpoint is exceeded, then the operator SHALL determine the weekly release rate and take actions to ensure the weekly release rate action level is not exceeded.
- 2. If the weekly release rate is determined to have been exceeded, then the reactor SHALL be placed in SHUTDOWN until the condition can be evaluated and the release rates determined to be below action levels.
3.7.5 EFFLUENTS - STACK MONITOR OPERABILITY The stack gaseous and particulate activity monitors SHALL be OPERATING when the reactor is operated above 0.1 kW or when any activity is performed in the facility that could release airborne radioactivity in the reactor cell. If either monitor is not functional:
- 1. Reduce power to below 0.1 kW
- 2. All evolutions that could precipitate airborne releases SHOULD be discontinued within the FACILITY.
- 3. The failed monitor SHOULD be restored to functionality by the end of the run or at the discretion of management.
- 4. If these actions cannot be completed, the reactor SHALL be placed in REACTOR SHUTDOWN and not returned to operation above 0.1 kW until both monitors are functional.
Proposed TS 3.7, Radiation and Environmental Monitoring Systems, states:
3.7.1 AREA RADIATION MONITOR An operational area radiation monitor* is required in the reactor cell and will alarm at 10 mr/hr or less.
- An operational area radiation monitor SHALL include:
Instrument readout that is visible in the control room a gamma-sensitive instrument A local audible alarm Alarm indication at a remote monitoring location 3.7.2 ENVIRONMENTAL MONITORING The VNC SITE utilizes environmental air sampling stations and thermoluminescent (TLD) and/or optically stimulated luminescence (OSL) dosimeters in locations specified by the VNC Environmental Monitoring Manual.
3.7.3 STACK MONITOR OPERABILITY The stack particulate activity monitor SHALL be OPERATING when any evolution is performed in the FACILITY that could generate an airborne concentration greater than one DAC in the reactor cell and will alarm at 2.0E-9 µCi/cc or less. If the monitor is not OPERABLE, all evolutions that could cause such airborne releases SHALL be discontinued within the FACILITY and corrective action taken to restore functionality.
In the Roadmap, section C, item 9, Section 3.5, Reactor Cell, Ventilation, and Confinement System, the licensee states that current TS 3.7, Radiation Monitoring Systems and Effluents, was proposed to be retitled to Radiation and Environmental Monitoring Systems to reflect that the facility is no longer generating radioactive effluents from reactor operations. The licensee also states that:
Current TS 3.7.1 was proposed to be deleted, as it is not applicable to the possession only status, and current TS 3.7.2 was proposed to be renumbered and retitled as TS 3.7.1, Area Radiation Monitor. The licensee proposed to add a setpoint and remote indication to enhance its radiation monitoring.
Current TS 3.7.3 was proposed to be renumbered and retitled as TS 3.7.2, Environmental Monitoring, and optically stimulated luminescence (OSL) dosimeters were proposed to be added as an equivalent means of measuring environmental radiation exposure.
Current TS 3.7.4 was proposed to be deleted as action levels based on operational effluent releases are not applicable in the possession-only status.
Current TS 3.7.5 was proposed to be renumbered and retitled as TS 3.7.3, Stack Monitor Operability; effluents was removed from the title as effluents from reactor operation are not produced in the possession-only status. The licensee proposed to revise proposed TS 3.7.3 to only monitor particulate activity, which is consistent with the possession-only status, and to add an alarm setpoint of 2.0E-9 microcuries per cubic centimeter (µCi/cc).
In addition, the licensee states that these proposed changes reflect the possession-only status of the facility, with reactor operations not authorized, and that the generation of radioactive effluents and fission products will essentially cease. The licensee indicated that the OSL dosimeters are used along with the thermoluminescent dosimeters (TLDs).
The NRC staffs review of proposed TS 3.7 finds that the licensee has maintained requirements for area radiation monitoring, environmental monitoring, and airborne release monitoring (stack monitor). The NRC staff finds that the area radiation monitor in proposed TS 3.7.1 is generally unchanged but now includes an alarm at 10 millirem per hour (mr/hr) or less, and an alarm indication at a remote location, which helps to ensure that any high radiation conditions will be available to the licensee for corrective actions. Further, the NRC staff finds that the setpoint of 10 mr/hr or less is acceptable as the licensee indicated that it is below background radiation readings, and that any increase will be identified. The NRC staff finds that the change to add OSL dosimeters to proposed TS 3.7.2 is consistent with the current facility practice for environmental radiation monitoring. The NRC staff finds that the changes to proposed TS 3.7.3 reflect the possession-only status of the NTR, and the elimination of reactor operation-generated effluents. However, since the licensee indicates that the primary coolant continuously vents into the reactor cell, any increase will be detectable by the Stack Monitor. The NRC staff finds that the licensees proposed setpoint of 2.0E-9 uCi/cc for the Stack Monitor releases is acceptable because the licensee indicates that it is slightly above background radiation readings, so that any minor increase in radiation will be identified.
Based on the above, the NRC staff concludes that proposed TS 3.7 is acceptable.
Current TS 3.8, Experiments, states:
3.8.1 EXPERIMENT REACTIVITY WORTH LIMIT The sum of the REACTIVITY WORTH of all EXPERIMENTS performed at any one time SHALL be limited to comply with the specification on POTENTIAL EXCESS REACTIVITY (Refer to LCO 3.1.1.).
3.8.2 EXPERIMENTAL OBJECT MOVEMENT No experimental object SHALL be moved during REACTOR OPERATION unless its potential REACTIVITY WORTH is known to be less than $0.50 3.8.3 EXPLOSIVES LIMITS FOR THE NTR The amounts of explosives (detonating and deflagrating, DOT Hazard Class/Divisions 1.1, 1.2, 1.3 and 1.4) permitted in the NTR facilities are as follows:
- i.
South Cell, W (D/2)2 with W 9 lbs and D 3 ft.
ii. North room (without Modular Stone Monument), W D22 with W 16 lbs and D 1ft.
iii. Setup Room, W 25 lbs.
3.8.4 EXPLOSIVES LIMITS FOR THE NORTH ROOM The amounts of explosives allowed in the North room MSM (inclusive in the limit of 3.8.3. ii. above) are as follows:
- i.
for DOT Hazard Class Divisions 1.1, 1.2, and 1.3 (detonating): W 2 pounds ii. for DOT Hazard Class Division 1.4 (deflagrating): W 4 pounds where: W = Total weight of explosives in pounds of equivalent TNT.
D = Distance in feet from the South Cell blast shield or the North Room wall.
3.8.5 EXPERIMENTAL OBJECTS IN THE CORE TANK Experimental objects SHALL not be allowed inside the core tank when the reactor is at a power greater than 0.1 kW.
3.8.6 EXPERIMENTAL OBJECTS IN THE FUEL LOADING CHUTE Experimental objects located in the fuel loading chute SHALL be secured to prevent their entry into the core region during REACTOR OPERATION.
3.8.7 RADIOACTIVE MATERIAL NEAR EXPLOSIVES A maximum of 10 Ci of radioactive material and up to 50 g of uranium SHALL be in storage in a neutron radiography area where explosive devices are present (i.e., in the South Cell or North Room). The storage locations SHALL be at least 1.5 m (5 ft) from any explosive device.
Radioactive materials, other than byproduct irradiated explosive devices and imaging systems, are not permitted in the Setup Room if EXPLOSIVE MATERIAL is present.
Exception. Devices containing not more than 10 grams TNT equivalent of explosives with up to 200 mCi of tritium in the form of tritiated metal (hydride) are permitted.
However, no more than one device SHALL be in a neutron radiography area or the setup room at any one time, and no other EXPLOSIVE MATERIAL SHALL be in the same area at that time.
3.8.8 EXPLOSIVES IN RADIATION FIELDS No explosive device SHALL be placed in a radiation field greater than 1 x 104 roentgens or consisting of greater than 3 x 1012 n/cm2 thermal neutrons.
3.8.9 ELECTROMAGNETIC WAVE NEAR EXPLOSIVES RESTRICTION With the exception of communication equipment utilizing low-energy electromagnetic waves in radiofrequencies, such as mobile phones and two-way hand-held radios, unshielded high-frequency generating equipment SHALL not be operated within 50 feet of any explosive device.
3.8.10 EXPERIMENTAL CAPSULE DESIGN Experimental capsules to be utilized in the EXPERIMENTAL FACILITIES SHALL be designed or tested to ensure that any pressure transient produced by chemical reaction of their contents and/or leakage of corrosion or FLAMMABLE materials will not damage the reactor.
3.8.11 FISSILE MATERIAL EXPERIMENTAL LIMITATIONS EXPERIMENTS containing fissile material SHALL be encapsulated and limited to a U-235 inventory of 50 mg.
3.8.12 CHEMICAL ENERGY FROM FLAMMABLE MATERIALS The potential REACTIVITY WORTH of any component which could be ejected from the reactor by a chemical reaction SHALL be less than $0.50 The maximum possible chemical energy release from the combustion of FLAMMABLE materials contained in any EXPERIMENTAL FACILITY SHALL not exceed 1000 kW-sec. The total possible energy release from chemical combination or decomposition of substances contained in any experimental capsule SHALL be limited to 5 kW-sec, if the rate of the reaction in the capsule could exceed 1 W.
EXPERIMENTAL FACILITIES containing FLAMMABLE materials SHALL be vented external to the reactor graphite pack.
3.8.13 EXPERIMENT APPROVAL A written description and analysis of the possible hazards involved for each type of EXPERIMENT SHALL be evaluated and approved by the area manager, or his designated alternate, before the EXPERIMENT is conducted.
3.8.14 EXPERIMENT INTERFERENCE IN REACTOR SHUTDOWN No irradiation SHALL be performed which could credibly interfere with the scram action of the SAFETY RODS at any time during REACTOR OPERATION.
3.8.15 EXPERIMENT RADIATION LIMITS The radioactive material content, including fission products, of any singly encapsulated EXPERIMENT to be utilized in the EXPERIMENTAL FACILITIES SHALL be limited, so that the complete release of all gaseous, particulate, or volatile components from the encapsulation could not result in doses in excess of 10% of the equivalent annual doses stated in 10 CFR Part 20. This dose limit applies to persons occupying unrestricted areas continuously for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> starting at time of release or restricted areas during the length of time required to evacuate the restricted area.
Proposed TS 3.8, Experiments, states:
3.8 EXPERIMENTS No specifications are applicable due to the reactor being in a POLSC.
Experiments are not authorized, and explosives are no longer stored in the FACILITY.
In the Roadmap, section C, item 10, Section 3.8, Experiments, the licensee states that the current TSs 3.8.1 through 3.8.15 were proposed to be deleted in their entirety since no experiments will be authorized in the possession-only status.
The NRC staffs review of proposed TS 3.8 finds that experiments are no longer authorized for the possession-only status of the NTR. The NRC staff also finds that deleting current TSs 3.8.1 through 3.8.15 is appropriate as experiments are only applicable to an operating reactor.
Further, the NRC staff finds proposed TS 3.8 consistent with the review guidance in NUREG-1537, Part 2, section 17.2.1.2. Therefore, the NRC staff concludes that proposed TS 3.8 is acceptable.
3.3.5 TS 4, Surveillance Requirements Current TS 4.0, General Surveillance Intervals, states:
4.0 GENERAL SURVEILLANCE INTERVALS Surveillances SHALL not exceed their defined SURVEILLANCE INTERVALS (Refer to Definitions, 1.2) unless deferred according to Surveillance Requirements 4.0.1 or 4.0.2.
4.0.1 DEFERRED OPERATING SURVEILLANCES Surveillances (except those required for safety while in REACTOR SHUTDOWN) MAY be deferred during a period which the reactor is shutdown, except, for Table 4-2 Items 2, 4, and 5 (Test and Calibration), and Surveillance Requirement 4.7.1 (Test and Calibration). Deferred surveillances SHALL be completed prior to reactor startup unless REACTOR OPERATION is required for performance of the surveillance. These surveillances SHALL be performed as soon as practical after startup.
4.0.2 DEFERRED SHUTDOWN SURVEILLANCES Scheduled surveillances which cannot be performed with the REACTOR OPERATING, MAY be deferred until the subsequent scheduled REACTOR SHUTDOWN.
In the Roadmap, section C, item 11, Section 4.0, General Surveillance Intervals, the licensee states that current TS 4.0 was applicable to an operating reactor and, therefore, proposed to delete it in its entirety. The NRC staffs review finds that current TS 4.0 is not applicable to the possession-only status of the NTR. Therefore, the NRC staff concludes that the proposed deletion of current TS 4.0 in its entirely is acceptable.
Current TS 4.1, Reactor Core Parameters, states:
4.1.1 POTENTIAL EXCESS REACTIVITY POTENTIAL EXCESS REACTIVITY SHALL be calculated before each startup.
Actual critical rod position SHALL then be used to verify that the MEASURED VALUE is $0.76.
4.1.2 SUBCRITICAL ROD POSITION The reactor SHALL be placed in REACTOR SHUTDOWN if it is not in a subcritical condition with all four SAFETY RODS withdrawn and all CONTROL RODS inserted during every reactor startup. SAFETY ROD withdrawal SHALL be stopped if it appears criticality will be reached before all SAFETY RODS are withdrawn.
4.1.3 MINIMUM SHUTDOWN MARGIN The minimum SHUTDOWN MARGIN SHALL be determined by calculation or measurement biennially or whenever a decrease in the reactivity worth of a SAFETY ROD is suspected.
Proposed TS 4.1, Reactor Core Parameters, states:
4.1 REACTOR CORE PARAMETERS No specifications are applicable due to the reactor being in a POLSC.
Reactor operations are not authorized. Fuel handling in support of defueling is the only activity allowed.
In the Roadmap, section C, item 12, Section 4.1, Reactor Core Parameters, the licensee states that current TS 4.1 was no longer applicable for the possession-only status, reactor operations are not authorized, and fuel handling in support of defueling is the only activity allowed for the reactor.
The NRC staffs review of proposed TS 4.1 finds that the reactor core parameters in current TS 4.1 are not applicable to the possession-only status of the NTR. The NRC staff also finds the deletion of current TS 4.1 consistent with the review guidance in NUREG-1537, Part 2, section 17.2.1.2. Therefore, the NRC staff concludes that the proposed deletion of current TS 4.1 and proposed TS 4.1 are acceptable.
Current TS 4.2, Reactor Control and Safety System, states:
4.2.1 RODS OPERABLE Each SAFETY ROD and CONTROL ROD drive SHALL be tested for operability annually.
4.2.2 SAFETY ROD WITHDRAWAL The interlock which restricts SAFETY ROD withdrawal to one rod at a time, in the pre-determined sequence, SHALL be tested annually.
4.2.3 SAFETY ROD WITHDRAWAL RATE The rate of withdrawal of each SAFETY ROD SHALL be measured annually.
4.2.4 CONTROL ROD WITHDRAWAL RATE The rate of withdrawal of each CONTROL ROD SHALL be measured annually.
4.2.5 SCRAM TIME The SAFETY ROD SCRAM TIME SHALL be measured semi-annually. The SCRAM TIME SHALL also be measured after any work is performed which could affect it.
4.2.6 REACTOR SAFETY SYSTEM AND SAFETY-RELATED ITEMS Checks, tests and calibrations of the REACTOR SAFETY SYSTEM and safety-related items SHALL be performed as specified in Tables 4-1 and 4-2 of these Technical Specifications.
Table 4-1 SURVEILLANCE REQUIREMENTS OF REACTOR SAFETY SYSTEM SCRAM INSTRUMENTS Item No.
System Surveillance Frequency*
- 1.
Linear Power CHANNEL CHECK (neutron source check)
Prior to SU CHANNEL TEST (high level trip test)
Prior to SU CHANNEL TEST (lack of high voltage)
Monthly CHANNEL CHECK (comparison against a heat balance)
Monthly CHANNEL CALIBRATION Annual
- 2.
Log N CHANNEL CHECK Prior to SU CHANNEL TEST Monthly CHANNEL CALIBRATION Annually
- 3.
Primary Coolant Temperature (Fenwall)
CHANNEL TEST Prior CHANNEL CALIBRATION Annually
- 4.
Primary Coolant Flow CHANNEL CHECK Prior to SU CHANNEL TEST Prior to SU CHANNEL CALIBRATION Annually
- 5.
Manual CHANNEL TEST Prior to SU
- 6.
Electrical Power CHANNEL TEST Prior to SU
- Prior to placing into service an instrument which has been repaired or declared INOPERABLE, the instrument check, or test or calibration, as appropriate will be performed to demonstrate operability.
Table 4-2 SURVEILLANCE REQUIREMENTS OF REACTOR SAFETY-RELATED ITEMS (INFORMATION INSTRUMENTS)
Item No.
System Surveillance Frequency*
- 1.
Reactor Cell Pressure CHANNEL CHECK Prior to SU CHANNEL TEST Quarterly CHANNEL CALIBRATION Annually
- 2.
Fuel Loading Tank Water Level CHANNEL TEST Quarterly
- 3.
Primary Coolant Temperature (TC-7)
CHANNEL TEST Quarterly CHANNEL CALIBRATION Annually
- 4.
Primary Coolant Temperatures (TC2 &
TC5)
CHANNEL CHECK Monthly CHANNEL CALIBRATION Annually
- 5.
Stack Radioactivity (Gas and particulate CHANNELS)
CHANNEL CHECK Prior to SU CHANNEL TEST Monthly CHANNEL CALIBRATION Annually
- 6.
Linear Power - Low Power Rod Block Setpoint CHANNEL TEST Monthly
- 7.
CONTROL ROD or SAFETY ROD not IN CHANNEL TEST Annually
- 8.
SAFETY ROD Sequence CHANNEL TEST Annually
- 9.
Primary Coolant Conductivity CHANNEL CHECK Quarterly CHANNEL CALIBRATION Biennially
- Prior to placing into service an instrument which has been repaired or declared INOPERABLE, the instrument check, or test, or calibration, as appropriate will be performed to demonstrate operability.
Proposed TS 4.2, Reactor Control and Safety System, states:
4.2.1 RODS INOPERABLE SAFETY RODS and CONTROL RODS SHALL be verified semi-annually to meet the conditions of the POLSC.
4.2.2 MANUAL POISON SHEETS SECURED MANUAL POISON SHEET covers SHALL be verified semi-annually to be locked in place and the keys removed from the FACILITY.
In the Roadmap, section C, item 13, Section 4.2, Reactor Control and Safety System, the licensee states that current TS 4.2.1, Rods Operable, was proposed to be retitled Rods Inoperable and was proposed to be changed to establish a semi-annual verification that both the safety and control rods were maintained consistent with the requirements of the POLSC as stated in proposed TS 5.3.1. Current TS 4.2.2 was proposed to be replaced by proposed TS 4.2.2, which requires a semi-annual verification that the MPS are locked in place and the keys removed from the facility. Current TSs 4.2.3 through 4.2.6 and TS tables 4-1 and 4-2 were proposed to be deleted in their entirety as they are no longer applicable to the possession-only status of the facility.
The NRC staffs review finds that proposed TSs 4.2.1 and 4.2.2 provide surveillance requirements to help ensure that the safety and control rods and the MPS continue to be maintained in the POLSC positions required by proposed TS 5.3.1. The NRC staff evaluated proposed TS 5.3.1 and finds the proposed TS acceptable (see SE section 3.3.6). The NRC staff also finds proposed TS 4.2 consistent with the review guidance in NUREG-1537, Part 2, section 17.2.1.2. Further, the NRC staffs review finds that the deletion of current TS 4.2 is appropriate as it is not applicable to the possession-only status of the NTR. Therefore, the NRC staff concludes that proposed TS 4.2 is acceptable.
Current TS 4.3, Reactor Coolant System, states:
Specifications regarding surveillance requirements of the reactor coolant system for flow, fuel loading tank level, and conductivity are included in the REACTOR SAFETY SYSTEM, Surveillance Requirements Section 4.2, Tables 4-1 and 4-2.
Proposed TS 4.3, Primary Coolant System, states:
4.3.1 FORCED FLOW COOLING The primary coolant flow instrument CHANNEL CHECK SHALL be performed quarterly and a CHANNEL CALIBRATION annually.
4.3.2 FUEL LOADING TANK FULL The fuel loading tank level SHALL be visually checked quarterly.
4.3.3 FUEL LOADING TANK LEVEL ALARM The fuel loading tank low level alarm CHANNEL TEST SHALL be performed quarterly.
4.3.4 PRIMARY COOLANT CONDUCTIVITY The primary coolant conductivity instrument CHANNEL CHECK SHALL be performed quarterly and a CHANNEL CALIBRATION biennially.
In the LAR, enclosure 2, section D, Primary Coolant System, the licensee indicates that the possession-only status includes the operation of the primary coolant system for chemistry control for the protection of the fuel cladding. In the Roadmap, section C, item 14, Section 4.3, Reactor Coolant System, the licensee states that the surveillances in proposed TS 4.3.1, TS 4.3.2, TS 4.3.3, and TS 4.3.4 are to help ensure that the primary coolant system is maintained operational, consistent with the requirements of the POLSC as stated in proposed TS 5.3.1. The licensee indicates that it proposed to change the title to Primary Coolant System to align with terminology used in proposed TS 3.3 and TS 3.5. The licensee also states that:
Proposed TS 4.3.1 was proposed to be added in support of proposed TS 3.3.1 and the surveillance intervals were proposed to be revised to perform quarterly channel checks and annual channel calibrations.
Proposed TS 4.3.2 was proposed to be added in support of proposed TS 3.3.2 to help ensure that the core tank remains full of water.
Proposed TS 4.3.3 was proposed to be added in support of proposed TS 3.3.3 to perform quarterly channel checks for the fuel loading tank level alarm.
Proposed TS 4.3.4 was proposed to be added in support of proposed TS 3.3.4 to perform quarterly channel checks and biennial calibrations of the primary coolant conductivity instrument.
The NRC staffs review finds that proposed TS 4.3.1, TS 4.3.2, TS 4.3.3, and TS 4.3.4 help ensure that the primary coolant system remains operational in support of the POLSC as required by proposed TS 5.3.1. The NRC staff evaluated proposed TS 5.3.1 and finds the proposed TS acceptable (see SE section 3.3.6). The NRC staff also finds that the proposed surveillance intervals are consistent with the review guidance in NUREG-1537, Part 2, section 17.2.1.2. Therefore, the NRC staff concludes that proposed TS 4.3 is acceptable.
Current TS 4.5, Reactor Cell Ventilation and Confinement System, states:
4.5.1 REACTOR CELL NEGATIVE PRESSURE Surveillance requirements for the instrumentation and equipment required to comply with LCO 3.5.1 SHALL be tested as listed in Surveillance Requirements Section 4.2, Table 4-2, Item No. 1 & 5.
4.5.2 REACTOR CELL ACTIVITY RELEASE A CHANNEL CHECK SHALL be performed DAILY during activities that could release airborne radioactivity into the reactor cell.
Proposed TS 4.5, Reactor Cell Ventilation, and Confinement System, states:
4.5.1 REACTOR CELL NEGATIVE PRESSURE The reactor cell differential pressure instrument CHANNEL CHECK SHALL be performed daily when the ventilation system is operating and a CHANNEL CALIBRATION annually.
In the Roadmap, section C, item 15, Section 4.5, Reactor Cell Ventilation and Confinement System, the licensee states that the surveillances in proposed TS 4.5 were proposed to be revised to reflect the need to continue to maintain the ventilation system during the possession-only status. The surveillances in proposed TS 5.4.1 were previously in TS table 4-2, item no. 1 and item no. 5.
The NRC staffs review finds the proposed TS 4.5 surveillance frequencies sufficient to help ensure the operability of the ventilation system when needed to help control airborne radioactivity in the reactor cell. The NRC staff also finds proposed TS 4.5 consistent with the review guidance in NUREG-1537, Part 2, section 17.2.1.2. Therefore, the NRC staff concludes that proposed TS 4.5 is acceptable.
Current TS 4.7, Radiation Monitoring Systems and Effluents, states:
4.7.1 MONITORING SYSTEMS DURING REACTOR OPERATIONS Surveillances for the Area Radiation Monitors during REACTOR OPERATIONS include a PRIOR to SU CHANNEL CHECK, a MONTHLY CHANNEL TEST, and an ANNUAL CHANNEL CALIBRATION. Prior to placing into service an Area Radiation Monitor which has been repaired or declared INOPERABLE, the applicable surveillance will be performed to demonstrate it is OPERABLE.
4.7.2 MONITORING SYSTEMS DURING REACTOR CELL MAINTENANCE A CHANNEL CHECK SHALL be performed DAILY during reactor cell maintenance.
4.7.3 EFFLUENTS - ENVIRONMENTAL MONITORING
- a. Monitoring of dose on SITE using thermoluminescent dosimeters or other equivalent devices SHALL be performed and documented annually.
- b. Environmental monitoring (e.g., sampling of soil and vegetation) SHALL be performed and documented annually.
4.7.4 EFFLUENTS - STACK RELEASE ACTIVITY The stack alarm SHALL be verified MONTHLY.
4.7.5 EFFLUENTS - STACK MONITOR OPERABILITY Stack activity monitors SHALL be performed according to Table 4-2, Item No. 5.
Proposed TS 4.7, Radiation and Environmental Monitoring Systems, states:
4.7.1 AREA RADIATION MONITOR Area Radiation monitor CHANNEL CHECK SHALL be performed quarterly and a CHANNEL CALIBRATION annually.
4.7.2 ENVIRONMENTAL MONITORING
- a. Monitoring of dose on SITE using thermoluminescent (TLD) and/or optically stimulated luminescence (OSL) dosimeters SHALL be performed and documented annually.
- b. Environmental monitoring (e.g., sampling of soil and vegetation) SHALL be performed and documented annually.
4.7.3 STACK MONITOR OPERABILITY Stack particulate activity monitor CHANNEL CHECK SHALL be performed daily when ventilation is required to be operated, and a CHANNEL CALIBRATION annually.
In the Roadmap, section C, item 16, Section 4.7, Radiation Monitoring Systems and Effluents, the licensee states that the title was proposed to be revised to reflect the possession-only status of the facility and that radioactive effluents would no longer be generated by reactor operations.
The licensee indicates that surveillances in current TS 4.7 were proposed to be revised to reflect the need to continue to maintain the reactor cell area radiation monitor, environmental monitoring, and stack particulate monitoring systems during the possession-only status. The surveillances in proposed TS 4.7.1 include revising the channel check from prior to reactor startup to quarterly for consistency with the possession-only status. The annual calibration would be unchanged. The annual surveillances proposed in TS 4.7.2 would be unchanged. The surveillances in proposed TS 4.7.3 include revising the channel check from prior to reactor startup to quarterly for consistency with the possession-only status. The annual calibration would be unchanged.
The NRC staffs review finds the proposed changes to current TS 4.7 consistent with the existing surveillance frequencies and sufficient to help maintain the radiation and environmental monitoring during the possession-only status of the NTR. The NRC staff also finds proposed TS 4.7 consistent with the review guidance in NUREG-1537, Part 2, section 17.2.1.2. Therefore, the NRC staff concludes that proposed TS 4.7 is acceptable.
Current TS 4.8, Experiments, states:
Specific surveillance activities SHALL be established during the review and approval process as specified in Administrative Control 6.2.3 "Review Function" and are not part of the Technical Specifications.
Proposed TS 4.8, Experiments, states:
These specifications are not applicable due to the reactor being in a POLSC.
Experiments are not authorized.
In the Roadmap, section C, item 17, Section 4.8, Experiments, the licensee states that the surveillances for experiments are not applicable in the POLSC status and that experiments and storage of explosive devices are no longer authorized at the facility.
The NRC staffs review finds the proposed changes to current TS 4.8 consistent with the possession-only status of the NTR. The NRC staff also finds proposed TS 4.8 consistent with the review guidance in NUREG-1537, Part 2, section 17.2.1.2. Therefore, the NRC staff concludes that proposed TS 4.8 is acceptable.
3.3.6 TS 5, Design Features Current TS 5.1.3, Effluent Discharge, states:
The discharge of all gaseous radioactive effluents SHALL be from the effluent stack at a minimum height of 45 feet (14 meters) above the grade level of Building 105.
The licensee proposes to delete current TS 5.1.3.
In the Roadmap, section C, item 18, Section 5.1, Site and Facility Description, the licensee proposes to delete current TS 5.1.3. The LAR safety analysis states that radioactive effluent from activation and fission products is no longer generated at the NTR in the possession-only status and that adequate dispersion remains in proposed TS 3.5.1 (i.e., reactor cell negative pressure) and is now considered a function of site radiation safety and not a design feature of the NTR.
The NRC staffs review finds the proposed deletion of current TS 5.1.3 appropriate as the generation of any radioactive effluent from activation and fission products will not occur as operation of the reactor will not be authorized. The radiation monitoring requirements in proposed TS 3.5, TS 3.7, TS 4.5, and TS 4.7 will ensure that effective radiation monitoring is provided during the possession-only status of the NTR. Therefore, the NRC staff concludes that the deletion of current TS 5.1.3 is acceptable.
Current TS 5.2, Reactor Primary Coolant System, states:
5.2.1 Primary System Pressure The reactor coolant system is maintained at atmospheric pressure by a vent line to the holdup tank and the top of the fuel tank being open to the reactor cell.
Proposed TS 5.2, Primary Coolant System, states:
5.2.1 Primary System Pressure The primary coolant system is maintained at atmospheric pressure by a vent line to the holdup tank and the top of the fuel tank being open to the reactor cell.
In the Roadmap, section C, item 19, Section 5.2, Reactor Primary Coolant System, the licensee states that reactor was proposed to be removed from the title of TS 5.2 and from TS 5.2.1 and replaced by primary to align with proposed TS 3.3 and TS 4.3.
The NRC staffs review finds the proposed changes consistent with proposed TS 3.3 and TS 4.3, as the primary coolant system will be maintained to control conductivity and, therefore, provide corrosion control for the fuel. Therefore, the NRC staff concludes that proposed TS 5.2 is acceptable.
Current TS 5.3, Reactor Core and Fuel, states:
5.3.1 CONTROL SYSTEM The control system SHALL consist of four scrammable, spring-actuated SAFETY RODS, three nonscrammable CONTROL RODS, and MANUAL POISON SHEETS.
Up to three MANUAL POISON SHEETS MAY be added or removed as needed to limit positive excess reactivity and compensate for reactivity loss from fuel burnup.
(1) The SAFETY RODS and CONTROL RODS SHALL be boron carbide clad in stainless steel.
(2) The MANUAL POISON SHEETS SHALL contain metallic cadmium.
(3) Each installed MANUAL POISON SHEET SHALL be restrained in its respective graphite reflector slot in a manner which will prevent movement by more than 1/2 inch relative to the reactor core.
(4) When the CONTROL RODS, SAFETY RODS, and MANUAL POISON SHEETS are inserted, they SHALL be located in the graphite reflector at the outer periphery of the core tank.
5.3.2 REACTOR FUEL The core SHALL consist of 16 fuel element assemblies. Each fuel element assembly SHALL consist of 40 disks separated by spacers of varying widths on an aluminum support shaft. Other nominal specifications of the assemblies SHALL include the following:
Fuel 23.5% (by weight uranium) / 76.5% aluminum (by weight aluminum)
Enrichment Approximately 93% U-235 (unburned)
Cladding Aluminum, 0.027-inch thickness Fuel disk active diameter 2.75 inch (OD)
Fuel disk spacing on shaft 0.24 to 0.27-inch, face-to-face 5.3.3 CORE REEL ASSEMBLY The fuel assemblies SHALL be positioned in a reel assembly inside the core tank.
The core reel assembly SHALL be rotated only when in REACTOR SHUTDOWN and by manual operation of a crank inside the NTR cell.
5.3.4 TEMPERATURE COEFFICIENT OF REACTIVITY The core is designed to exhibit a negative temperature coefficient of reactivity above 124°F, which is approximately the reactor steady-state operating temperature.
Proposed TS 5.3, Reactor Core and Fuel, states:
5.3.1 POSSESSION ONLY LICENSE SHUTDOWN CONFIGURATION (POLSC)
SAFETY and CONTROL RODS Remain fully inserted and restrained from any movement by ensuring:
SAFETY ROD drive belts removed.
SAFETY RODS electrically isolated.
Course CONTROL ROD drive chains removed.
CONTROL RODS electrically isolated.
MPS MPS of sizes 0.9, 0.5, and a full width poison sheet are installed and latched in slots #1, 2, and 5, respectively.
MPS covers installed, locked and the keys removed from the FACILITY.
Primary coolant Primary coolant system is OPERABLE.
Fuel loading tank is filled with water.
5.3.2 Core Reel Assembly The fuel assemblies SHALL be positioned in a reel assembly inside the core tank.
The core reel assembly SHALL be rotated only during authorized fuel handling activities and by manual operation of a crank inside the NTR reactor cell.
5.3.3 Temperature Coefficient of Reactivity The core is designed to exhibit a negative temperature coefficient of reactivity above 124°F, which is approximately the reactor steady-state operating temperature.
In the LAR, enclosure 2, section D, Control & Safety Rods, and in the Roadmap, section C, item 20, Section 5.3, Reactor Core and Fuel, the licensee states that current TS 5.3.1 was proposed to be renamed Possession Only License Shutdown Configuration (POLSC) and be revised with requirements for the control and safety rods, the MPS, and the primary coolant system consistent with the possession-only status of the facility. The licensee states that the sizes and locations of the MPS are consistent with the LAR safety analysis. The licensee included the primary coolant system in TS 5.3.1 for completeness as the requirements for the operation of the system are also included in proposed TS 3.3.
In the Roadmap, section C, item 20, the licensee also provided illustrations of the control and safety rods drive mechanisms to demonstrate that electrically isolating and removing the drive belts disables the rods and prevents any movement.
The licensee also proposed to delete current TS 5.3.2, Reactor Fuel, since it supported refueling operation of the NTR, which will not be authorized. The licensee proposed to renumber current TS 5.3.3 to proposed TS 5.3.2 and proposed to change the specification for the rotation of the core reel assembly from reactor shutdown to during authorized fuel handling activities.
The licensee proposed to renumber current TS 5.3.4 to proposed TS 5.3.3 due to the proposed deletion of current TS 5.3.2.
The NRC staffs review finds that the requirements of proposed TS 5.3.1 are consistent with the configuration of the CSA (baseline case), which the licensee used to justify that the POLSC would remain subcritical during all perturbations and configuration changes as analyzed in the CSA. The CSA was reviewed and found acceptable by the NRC staff in SE section 3.5. The NRC staffs review finds proposed TS 5.3.2 consistent with the possession-only status of the NTR and that the fuel will not be moved until final disposal. Therefore, the NRC staff concludes that proposed TS 5.3 is acceptable.
Current TS 5.4, Fissionable Material Storage, states:
5.4.1 FUEL STORAGE Fuel including fueled EXPERIMENTS and fuel devices not in the reactor SHALL be stored in a geometrical array where keff is no greater than 0.9 for all conditions of moderation and reflection using light water.
The licensee proposes to delete current TS 5.4.
In the LAR, section C, Hazards, the licensee states that all SNM and non-reactor byproduct material has been removed from the NTR facility except for the fuel in storage in the reactor. In the Roadmap, section A, Storage of Fuel, the licensee indicated that the fuel would need to remain in storage in the reactor because there were no other storage location options available.
The NRC staffs review finds that proposed TS 5.3.2 requires that the fuel assemblies only be stored in the reel assembly inside the core tank, and no other storage locations are authorized.
The NRC staff finds that the requirements in current TS 5.4 are no longer applicable to the NTR facility. Therefore, the NRC staff concludes that the proposed deletion of current TS 5.4 is acceptable.
3.3.7 TS 6, Administrative Controls Current TS 6.1, Organization, states:
The NTR SHALL be owned and operated by the LICENSEE with management and operations organization as shown in Figure 6-1.
6.1.1 STRUCTURE Figure 6-1 FACILITY Organization 6.1.2 RESPONSIBILITIES (1) The Level 1 manager SHALL be responsible for the NTR FACILITY LICENSE.
(2) The Level 2 manager is designated the area manager for the NTR and SHALL be responsible for the overall safe operation and maintenance of the FACILITY.
(3) The Level 3 Reactor supervisor (if utilized) is the individual responsible for supervising daily operations. In the absence of this position, the Level 2 manager is responsible for supervising daily operations.
(4) The Level 4 Operations staff includes SENIOR REACTOR OPERATORS, REACTOR OPERATORS, and trainees.
(5) Responsibilities of one level MAY be assumed by alternates when designated in writing.
(6) Functions performed by one level MAY be performed by a higher level, provided the minimum qualifications are met (e.g., SENIOR REACTOR OPERATOR LICENSE).
6.1.3 STAFFING (1) The minimum staffing when the REACTOR IS NOT SECURED (Refer to REACTOR SECURED.) SHALL be composed of:
A LICENSED REACTOR OPERATOR in the control room.
A second person present at the SITE who is familiar with the VNC Radiological Emergency Plan and Emergency Procedures relevant to the NTR and is capable of carrying out FACILITY written procedures.
A LICENSED SENIOR REACTOR OPERATOR SHALL be present at the NTR FACILITY, or a READILY AVAILABLE SENIOR REACTOR OPERATOR designated.
(2) A list of reactor FACILITY personnel by name and telephone number SHALL be available in the control room for use by the operator and includes:
Management personnel Radiation safety personnel Other operations personnel (3) A LICENSED SENIOR REACTOR OPERATOR SHALL be present at the NTR FACILITY during the following events:
first daily startup and approach to power recovery from an UNSCHEDULED SHUTDOWN all reactor fuel, SAFETY ROD, and CONTROL ROD relocations within the reactor core region MANUAL POISON SHEET changes relocation of any EXPERIMENT or FACILITY changes with a REACTIVITY WORTH greater than one dollar.
6.1.4 SELECTION AND TRAINING OF PERSONNEL The selection, training and requalification of operations personnel SHALL meet or exceed the requirements of American National Standard for Selection and Training of Personnel for Research Reactors, ANSI/ANS 15.4-2016, and the latest revision of the FACILITY Operator Requalification Program.
Proposed TS 6.1, Organization, states:
The NTR SHALL be owned and maintained by the LICENSEE with the management and support organization as shown in Figure 6-1.
6.1.1 STRUCTURE Figure 6-1 FACILITY Organization 6.1.2 RESPONSIBILITIES (1) The Level 1 LICENSE Holder SHALL be responsible for the NTR FACILITY LICENSE.
(2) The Level 2 Reactor Administrator SHALL ensure that the NTR is maintained according to the FACILITY LICENSE and applicable regulations, and is responsible for security and safety of the FACILITY.
(3) Radiation Safety Function - Radiation Safety Officer is a SITE-wide compliance role that operates independently of the Reactor Administrator and is responsible for safe radiological practices and procedures on the SITE.
(4) Level 3 Certified Fuel Handler Supervisor is responsible for fuel handling operations and ensures the fuel handling operations are done safely, that staffing is adequate, and that Certified Fuel Handlers have current documented training and qualifications.
(5) The Level 4 Certified Fuel Handler performs fuel handling operations under the direction of the Certified Fuel Handler Supervisor.
(6) Responsibilities of one level MAY be assumed by alternates when designated in writing.
6.1.3 STAFFING The reactor SHALL not be operated and fuel movement prior to defueling is not permitted. A Certified Fuel Handler Supervisor SHALL be present in the reactor cell during fuel handling operations.
6.1.4 SELECTION AND TRAINING OF PERSONNEL (1) The Reactor Administrator SHALL meet minimal standards for this position that include a cumulative 5 years of reactor experience, with 2 years in an occupational radiation exposure program, and 2 years of personnel supervisory experience. Variations in these standards SHALL be justified in writing by the LICENSE Holder.
(2) Certified Fuel Handler Supervisors and Certified Fuel Handlers SHALL be trained in accordance with the NRC approved Certified Fuel Handler training program for the NTR.
In the Roadmap, section C, item 22, Section 6.1, Organization, the licensee describes the changes associated with proposed TS 6.1. In proposed TS 6.1, the licensee replaces operated with maintained and operations with support. In proposed TS 6.1.1, the licensee replaces the titles of Manager, Reactor Supervisor, and Operating Staff with License Holder, Reactor Administrator, Certified Fuel Handler Supervisor, and Certified Fuel Handler, respectively, to align with the possession-only status of the facility and the replacement of NRC-licensed operators with CFHs. In proposed TS 6.1.2, the licensee revises the titles to align with the new titles described in proposed TS 6.1.1 and adds a description of the Radiation Safety Function to align with proposed TS figure 6-1. In proposed TS 6.1.3, the licensee adds a requirement that fuel movement is restricted to defueling and requires a CFH Supervisor to be present in the reactor cell during fuel handling operations. In proposed TS 6.1.4, the licensee revises the requirements to align with the possession-only status of the facility by adding Reactor Administrator and CFH Supervisors and CFHs.
The NRC staffs review finds that the proposed changes to TS 6.1 are consistent with the possession-only status of the NTR. The NRC staff also finds that the organization structure in proposed TS 6.1.1 is consistent with the review guidance in NUREG-1537, Part 2, section 17.2.1.2 and ANSI/ANS-15.1-2007 (R2013), which states that the reporting and communication responsibilities should be clearly defined for the organizational levels (i.e., Level 1, 2, etc.), and that the oversight and radiation protection functions have been maintained. In addition, the NRC staff finds that the proposed changes to TSs 6.1.2, 6.1.3, and 6.1.4 reflect the possession-only status of the NTR, including new titles of the staff and the use of CFHs instead of NRC-licensed operators. Further, the NRC staff reviewed and found acceptable the licensees CFH training program in SE section 3.6. Based on the above, the NRC staff concludes that proposed TS 6.1 is acceptable.
Current TS 6.2, Review and Audit, states:
6.2.1 COMPOSITION AND QUALIFICATIONS (1) The RC organization SHALL conduct routine audits and perform periodic reviews of the implementation of these Technical Specifications.
(2) The Vallecitos Technological Safety Council (VTSC), at the direction of the Level 1 manager, SHALL perform independent reviews to ensure proper ongoing operation of the NTR.
(3) The VTSC SHALL not have more than half of its members from either Operations or RC Organizations.
(4) The VTSC SHALL be composed of a minimum of three members.
(5) VTSC members and alternates SHALL be appointed by the Level 1 manager.
(6) VTSC members SHALL collectively represent a broad spectrum of expertise in the appropriate reactor technology.
(7) Qualified and approved alternates MAY serve in the absence of regular members.
6.2.2 CHARTER AND RULES The VTSC functions SHALL be conducted under a written charter including provision for:
(1) A meeting frequency of not less than once per calendar year.
(2) Allowing only one vote for each member or alternate for each issue reviewed.
(3) Quorum rules whereby a quorum is at least one-half of the voting members, and the NTR operations staff doesnt constitute a majority of the quorum.
(4) The use of support organizations.
(5) Maintenance of records; including the dissemination, review, and approval of minutes.
6.2.3 REVIEW FUNCTION Activities requiring review SHALL include the following:
(1) Determinations that proposed changes in equipment, systems, tests, EXPERIMENTS, or procedures are allowed without prior NRC approval as determined by 50.59 evaluation.
(2) Determinations that new EXPERIMENTs or classes of EXPERIMENTs that could affect reactivity or result in the release of radioactivity do not require prior NRC approval as determined by 50.59 evaluation.
(3) Determinations that proposed changes to the Fire Protection program as described in the Safety Analysis Report that do not require prior NRC approval, would not adversely affect the ability to achieve and maintain safe REACTOR SHUTDOWN of the NTR in the event of a fire as determined by 50.59 evaluation.
(4) All new procedures and major revisions of existing procedures having safety significance that are required by the administrative control specifications in Administrative Controls Section 6.4.
(5) Proposed changes to the Technical Specifications or the FACILITY operating LICENSE.
(6) Violations of Technical Specifications, and FACILITY LICENSE requirements.
(7) Unusual or abnormal occurrences which are reportable to the NRC under provisions of the Federal Regulations or Administrative Control 6.7.2.
(8) Significant operating abnormalities or deviations from normal and expected performance of FACILITY equipment that affect, or could affect, nuclear safety.
(9) Audit Reports.
6.2.4 AUDIT FUNCTION Audits SHALL include examination of operations records, logs, and documents as well as discussions with staff and observations as appropriate. Deficiencies SHALL be reported to the Level 1 manager as soon as identified and a written report of the findings of the audit submitted to the Level 1 manager within 3 months after the audit has been completed. The following SHALL be audited:
(1) FACILITY operation for conformance to these Technical Specifications and applicable LICENSE conditions: at least once per calendar year not to exceed 15 months between audits.
(2) Retraining and requalification program for the LICENSED operations staff: at least once every other calendar year not to exceed 30 months between audits.
(3) The results of condition reports initiated relative to the NTR and operation of the NTR: once per calendar year not to exceed 15 months between audits.
(4) The VNC Radiological Emergency Plan and implementing procedures: once every other year not to exceed 30 months between audits.
Proposed TS 6.2, Review and Audit, states:
6.2.1 COMPOSITION AND QUALIFICATIONS (1) The Oversight Committee SHALL conduct routine audits and perform periodic reviews of the implementation of these Technical Specifications.
(2) The Oversight Committee SHALL be composed of the Level 2 Reactor Administrator and a member of radiation protection staff along with at least three individuals having expertise in reactor technology or radiation protection. Members SHALL be appointed by the Level 1 LICENSE Holder.
6.2.2 CHARTER AND RULES The Oversight Committee SHALL be conducted under a written charter including provisions for:
(1) A meeting frequency of not less than once per calendar year.
(2) Allowing only one vote for each member or alternate for each issue reviewed.
(3) Quorum rules whereby a quorum is at least one-half of the voting members.
(4) The use of support organizations.
(5) Maintenance of records; including the dissemination, review, and approval of minutes.
6.2.3 REVIEW FUNCTION Activities requiring review SHALL include the following:
(1) Determinations that proposed changes in equipment, systems, tests, or procedures are allowed without prior NRC approval as determined by 10 CFR 50.59 evaluation.
(2) All new procedures and major revisions of existing procedures having safety significance that are required by the administrative control specifications in Section 6.4.
(3) Proposed changes to the Technical Specifications or the FACILITY LICENSE.
(4) Violations of Technical Specifications, and FACILITY LICENSE requirements.
(5) Audit Reports.
6.2.4 AUDIT FUNCTION Audits SHALL include examination of operations records, logs, and documents as well as discussions with staff and observations as appropriate.
Deficiencies SHALL be reported to the Level 1 LICENSE Holder as soon as identified and a written report of the findings of the audit submitted to the Oversight Committee within 3 months after the audit has been completed.
The following SHALL be audited:
(1) FACILITY activities for conformance to these Technical Specifications and applicable LICENSE conditions: at least once per calendar year not to exceed 15 months between audits.
(2) Certified fuel handling training program: at least once every other calendar year not to exceed 30 months between audits.
(3) The results of condition reports initiated relative to the NTR: once per calendar year not to exceed 15 months between audits.
(4) NTR emergency response implementing procedures: once every other year not to exceed 30 months between audits.
In the Roadmap, section C, item 23, Section 6.2, Review and Audit, the licensee describes the changes to proposed TS 6.2. In proposed TS 6.2.1, the licensee adds the Oversight Committee in place of the Vallecitos Technical Safety Council (VTSC) and aligns the titles with proposed TS 6.1.1 for the possession-only status of the facility. In proposed TS 6.2.2, the licensee replaces the VTSC with the Oversight Committee and removes references to NTR operations staff. Other changes include: TS 6.2.3, item (2) removes reference to experiments as they are no longer authorized; TS 6.2.3, item (3) removes the review requirement for the fire protection program since the facility will be in a possession-only status; TS 6.2.3, item (7) removes the review requirement for internal reports to the NRC since they are contained in proposed TS 6.7.2; and TS 6.2.3, item (8) removes the review requirement for significant operating abnormalities since the reactor is no longer authorized to operate.
The NRC staffs review finds the proposed changes to current TS 6.2 consistent with the possession-only status of the NTR. The NRC staff also finds that the proposed TS 6.2 review and audit function remains consistent with the guidance in NUREG-1537, Part 2, section 17.2.1 and ANSI/ANS-15.1-2007 (R2013), which provides guidance for the composition and qualifications, charter and rules, and review and audit functions. In addition, the NRC staff finds that the proposed replacement of the VTSC with the Oversight Committee and the proposed removal of the review and audit function for those activities no longer authorized under the POL (e.g., experiments) are acceptable. Further, the NRC staff finds that the requirements in current TS 6.2.3, items (2), (3), (7), and (8) are no longer applicable to the possession-only status of the facility. Also, the NRC staff reviewed and found acceptable the licensees CFH training program in SE section 3.6. Based on the above, the NRC staff concludes that proposed TS 6.2 is acceptable.
Current TS 6.3, Radiation Safety, states:
The Level 2 manager (or the Level 3 supervisor when assigned), in coordination with the VNC Radiation Safety Officer (RSO), SHALL be responsible for implementing the NTR radiation safety function. The RSO SHALL report relevant findings to the Level 2 manager, but SHALL report organizationally to the Manager, RC, thereby maintaining independence from the reactor operations organization. The radiation safety function is informed by the guidelines of the ANSI/ANS 15.11-2016, Radiation Protection at Research Reactor Facilities.
Proposed TS 6.3, Radiation Safety, states:
The Level 2 Reactor Administrator (or the Level 3 Certified Fuel Handler Supervisor in his absence), in coordination with the VNC Radiation Safety Officer (RSO), SHALL be responsible for implementing the NTR radiation safety function. The RSO SHALL report relevant findings to the Level 2 Reactor Administrator but SHALL report organizationally to the Level 1 LICENSE Holder, thereby maintaining independence from the production organization.
The radiation safety function is informed by the guidelines of the ANSI/ANS 15.11-2016, Radiation Protection at Research Reactor Facilities.
In the Roadmap, section C, item 24, Section 6.3, Radiation Safety, the licensee states that the changes to proposed TS 6.3 are to align the organizational titles and responsibilities consistent with the organization structure in proposed TS 6.1 and that the change from operations to production is to align with the possession-only status of the facility.
The NRC staffs review finds that the proposed changes to TS 6.3 are consistent with the possession-only status of the NTR, the implementation of a CFH Training program, and the organizational titles in proposed TS 6.1. Therefore, the NRC staff concludes that proposed TS 6.3 is acceptable.
Current TS 6.4, Procedures, states:
Written procedures SHALL be prepared, reviewed, and authorized prior to initiating any of the activities listed in this section. Because the VNC is a multi-license FACILITY, procedures implementing elements of SITE-wide programs (i.e., radiation protection, emergency planning, security) are authorized by the SITE Manager, RC. NTR-specific implementing procedures as components of those larger programs SHALL be authorized by the Level 2 manager according to Administrative Control 6.4.2. Procedures exclusive to the implementation of administrative and operational requirements of the NTR Licensing basis and their revisions SHALL be authorized by the Level 2 manager or his designated alternate(s) according to this section. Several of the activities in Administrative Control 6.4.1 MAY be included in a single manual or set of procedures or divided among various manuals or procedures.
6.4.1 WRITTEN PROCEDURES Written procedures SHALL be prepared for the following activities as required:
(1) Startup, operation, and shutdown of the reactor.
(2) Defueling, refueling, and fuel transfer operations, when required.
(3) Preventive or corrective maintenance which could have an effect on the safety of the reactor, including the replacement of components.
(4) Surveillance checks, tests, calibrations, and inspections required by the Technical Specifications.
(5) NTR-specific radiation protection program implementing procedures for personnel safety consistent with applicable regulations or guidelines.
Management commitment and programs to maintain exposures and releases as low as reasonably achievable SHALL be a component of the SITE-wide radiation protection program.
(6) Administrative controls for operation and maintenance and the conduct of EXPERIMENTS that could affect reactor safety or core reactivity.
(7) NTR-specific implementing procedures for the SITE-wide emergency and security plans.
(8) NTR-specific radiation protection program implementing procedures for the use, receipt, and on-SITE transfer of by-product material for such activities performed under the R-33 LICENSE.
6.4.2 LEVEL 2 APPROVAL (1) The Level 2 manager SHALL authorize all new procedures required by Administrative Control 6.4.1 before implementation.
(2) The Level 2 manager SHALL authorize all non-ADMINISTRATIVE CHANGES to procedures required according to Administrative Control 6.4.1.
6.4.3 ADMINISTRATIVE CHANGES TO PROCEDURES (1) ADMINISTRATIVE CHANGES to procedures required by Administrative Control 6.4.1 MAY be made by the Level 3 reactor supervisor or Level 2 manager before implementation.
(2) ADMINISTRATIVE CHANGES made by authorization of the Level 3 reactor supervisor SHALL be subsequently approved by the Level 2 manager.
6.4.4 TEMPORARY DEVIATIONS Temporary deviations from established procedures MAY be made by a LICENSED SENIOR REACTOR OPERATOR in order to deal with special or unusual circumstances. These deviations SHALL be documented and reported to the Level 2 manager by the end of the next working day Proposed TS 6.4, Procedures, states:
Written procedures SHALL be prepared, reviewed, and authorized prior to initiating any of the activities listed in this section. Because the VNC is a multi-LICENSE FACILITY, procedures implementing elements of SITE-wide programs (i.e., radiation protection, emergency planning, security) are authorized by the Level 1 LICENSE Holder. NTR-specific implementing procedures as components of those larger programs and non-administrative changes to those procedures SHALL be authorized by the Level 2 Reactor Administrator. Procedures exclusive to the implementation of administrative requirements of the NTR Licensing basis and their revisions SHALL be authorized by the Level 2 Reactor Administrator or his designated alternate(s) according to this section. Several of the activities in Section 6.4.1 MAY be included in a single manual or set of procedures or divided among various manuals or procedures.
6.4.1 WRITTEN PROCEDURES Written procedures SHALL be prepared for the following activities as required:
(1) Fuel Handling - Defueling. These may be maintained inactive as fuel handling will not be performed under the POLSC prior to defueling.
(2) Preventive or corrective maintenance which could have an effect on the safety of the fuel in storage, including the replacement of components.
(3) Surveillance checks, tests, calibrations, and inspections required by the Technical Specifications.
(4) NTR-specific radiation protection program implementing procedures for personnel safety consistent with applicable regulations or guidelines.
Management commitment and programs to maintain exposures and releases as low as reasonably achievable SHALL be a component of the SITE-wide radiation protection program.
(5) NTR-specific implementing procedures for the SITE-wide emergency and security plans.
(6) NTR-specific radiation protection program implementing procedures for the use, receipt, and on-SITE transfer of by-product material for such activities performed under the R-33 LICENSE.
6.4.2 ADMINISTRATIVE CHANGES TO PROCEDURES (1) ADMINISTRATIVE CHANGES to procedures required by Section 6.4.1 MAY be made by the Level 3 Certified Fuel Handler Supervisor or Level 2 Reactor Administrator before implementation.
(2) ADMINISTRATIVE CHANGES made by authorization of the Level 3 Certified Fuel Handler Supervisor SHALL be subsequently approved by the Level 2 Reactor Administrator.
6.4.3 TEMPORARY DEVIATIONS Temporary deviations from established procedures MAY be made by a Level 3 Certified Fuel Handler Supervisor in order to deal with special or unusual circumstances. These deviations SHALL be documented and reported to the Level 2 Reactor Administrator by the end of the next working day.
In the Roadmap, section C, item 25, Section 6.4, Procedures, the licensee states that the proposed changes to TS 6.4 reflect the organizational titles and responsibilities consistent with the organization structure in proposed TS 6.1. Further, TS 6.4.1(1) was proposed to be deleted since reactor operation is not authorized in the possession-only status. TS 6.4.1(2) was proposed to be renumbered to TS 6.4.1(1) and to be revised to reflect fuel handling procedures consistent with the possession-only status. TS 6.4.3(3) was proposed to be revised to reflect the fuel in storage condition of the possession-only status. TS 6.4.1(6) was proposed to be deleted since experiments are not authorized in the possess-only status. The remaining TSs were proposed to be sequentially renumbered. TS 6.4.2, TS 6.4.3, and TS 6.4.4 were proposed to be revised to reflect the organizational titles and responsibilities consistent with the organizational structure in proposed TS 6.1.
The NRC staffs review finds that the proposed changes to TS 6.4 are consistent with the possession-only status of the NTR, procedures are related to fuel handling, and the titles reflect the organizational structure in proposed TS 6.1. Based on the above, the NRC staff concludes that proposed TS 6.4 is acceptable.
Current TS 6.5, Experiments Review and Approval, states:
6.5.1 NEW EXPERIMENT APPROVAL All new EXPERIMENTs or class of EXPERIMENTs SHALL undergo review according to Administrative Control 6.2.3 and be approved in writing by the Level 2 manager or designee.
6.5.2 CHANGES TO EXPERIMENTS Changes, except for ADMINISTRATIVE CHANGES, to EXPERIMENT implementing documents or to previously approved EXPERIMENTS SHALL undergo review according to Administrative Control 6.2.3 and be approved in writing by the Level 2 manager or designee.
6.5.3 ADMINISTRATIVE CHANGES TO EXPERIMENTS ADMINISTRATIVE CHANGES made to previously approved EXPERIMENT implementing procedures (e.g., ERs and EAFs) do not require independent review and MAY be approved by an SRO.
Proposed TS 6.5, Experiments Review and Approval, states:
Experiments are no longer performed at the NTR.
In the Roadmap, section C, item 26, Section 6.5, Experiments Review and Approval, the licensee states that the proposed changes to TS 6.5 reflect the possession-only status of the facility that does not authorize any experiments.
The NRC staffs review finds that the proposed changes to TS 6.5 are consistent with the possession-only status of the NTR that experiments are no longer authorized. Therefore, the NRC staff concludes that proposed TS 6.5 is acceptable.
Current TS 6.6, Required Actions, states:
6.6.1 Actions to be Taken in Case of Safety Limit Violation (1) The reactor SHALL be placed in REACTOR SHUTDOWN, and REACTOR OPERATIONs SHALL not be resumed until authorized by Level 1 management and the NRC.
(2) The safety limit violation SHALL be promptly reported to the Level 2 manager or designated alternates.
(3) The safety limit violation SHALL be reported to the NRC.
(4) A safety limit violation report SHALL be prepared. The report SHALL describe the following:
(a) Applicable circumstances leading to the violation including, when known, the cause and contributing factors.
(b) Effect of the violation upon reactor FACILITY components, systems, or structures and on the health and safety of personnel and the public.
(c) Corrective action to be taken to prevent recurrence.
(5) The report SHALL be reviewed by the Manager, Regulatory Compliance (RC) or designee and any follow-up report SHALL be submitted to the NRC when authorization is sought to resume operation of the reactor.
6.6.2 Action to be taken in the event of an occurrence of the type Identified in Section 6.7.2(1)b and 6.7.2(1)c (1) Reactor conditions SHALL be returned to normal or the reactor SHALL be placed in REACTOR SHUTDOWN. If REACTOR SHUTDOWN is necessary to correct the occurrence, operations SHALL not be resumed unless authorized by the Level 2 manager or the Level 1 manager.
(2) Occurrence SHALL be reported to the area manager and to the NRC addressed in accordance with 10 CFR 50.4.
(3) Occurrence SHALL be reviewed by the Manager, RC, or designee, or the VTSC at its next scheduled meeting.
Proposed TS 6.6, Required Actions, states:
Actions in response to safety limit violations are not applicable under the POLSC.
Reactor operations are not authorized. Internal reporting requirements are included in Section 6.7, Reports.
In the Roadmap, section C, item 27, Section 6.6, Required Actions, the licensee states that the content of current TS 6.6 was proposed to be deleted or relocated to proposed TS 6.7. Current TS 6.6 requires reporting violations of the Safety Limit, which is not applicable in the possession-only status. Current TS 6.6.2(1) requires the reactor to be shutdown, which is not applicable in the possession-only status.
The NRC staffs review finds that the proposed changes to TS 6.6 are consistent with the possession-only status of the NTR and that the proposed deletions are for requirements that involve a reactor authorized to operate. Therefore, the NRC staff concludes that proposed TS 6.6 is acceptable.
Current TS 6.7, Reports, states:
6.7.1 Operating Reports Annual operating report(s) SHALL be submitted to the NRC Document Control Desk.
The report(s) SHALL include the following:
(1) A narrative summary of reactor operating experience including the hours the reactor was critical and total energy produced.
(2) The UNSCHEDULED SHUTDOWNS including, where applicable, corrective action taken to preclude recurrence.
(3) Tabulation of major preventive and corrective maintenance operations having safety significance.
(4) A summary report in accordance with 10 CFR 50.59(d)(2).
(5) A summary of the nature and amount of radioactive effluents released or discharged to environs beyond the effective control of the owner-operator as determined at or before the point of such release or discharge. The summary SHALL include to the extent practicable an estimate of individual radionuclides present in the effluent. If the estimated average release after dilution or diffusion is <25% of the concentration allowed or recommended, a statement to this effect is sufficient.
(6) Summarized results of environmental surveys performed outside the FACILITY.
(7) A summary of exposures received by FACILITY personnel and visitors where such exposures are greater than 25% of that allowed or recommended.
6.7.2 Special Reports Special reports are used to report unplanned events as well as planned major FACILITY and administrative changes. The following special reports SHALL be forwarded to the NRC addressed in accordance with 10 CFR 50.4:
(1) There SHALL be a report not later than the following working day by telephone and confirmed in writing by telegraph or similar conveyance to the NRC, to be followed by a written report within 14 days, that describes the circumstances of any of the following events:
- a. Violation of safety limit
- b. Release of radioactivity from the SITE above allowed limits.
- c. Any of the following:
- i.
Operation with actual safety-system settings for required systems less conservative than the limiting safety-system settings specified in the Technical Specifications.
ii. Operation in violation of limiting conditions for operation established in the Technical Specifications unless prompt remedial action is taken.
iii. A REACTOR SAFETY SYSTEM component malfunction which renders or could render the REACTOR SAFETY SYSTEM incapable of performing its intended safety function unless the malfunction or condition is discovered during maintenance tests or REACTOR SHUTDOWN periods.
NOTE: Where components or systems are provided in addition to those required by the Technical Specifications, the failure of the extra components or systems are not considered reportable provided that the minimum numbers of components or systems specified or required perform their intended reactor safety function.
iv. An unanticipated or uncontrolled change in reactivity greater than $0.50.
- v. Abnormal and significant degradation in reactor fuel, cladding, or coolant boundary, which could result in exceeding prescribed radiation limits for personnel or the environment.
vi. An observed inadequacy in the implementation of administrative or procedural controls such that the inadequacy causes or could have caused the existence or development of an UNSAFE CONDITION with regard to REACTOR OPERATIONs.
(2) There SHALL be a written report within 30 days to the NRC for:
- a. Permanent changes in the FACILITY organization involving Level 1 or Level 2 management.
- b. Significant changes in the transient or accident analysis as described in the Safety Analysis Report.
Proposed TS 6.7, Reports, states:
6.7.1 ROUTINE REPORT A routine report providing the following information SHALL be submitted to the NRC Document Control Desk in accordance with the provisions of 10 CFR 50.59 not to exceed 24 months:
(1) Tabulation of major preventive and corrective maintenance activities having safety significance.
(2) A report in accordance with 10 CFR 50.59(d)(2) containing a brief description of any changes, tests, and experiments, including a summary of the evaluation of each.
(3) Summarized results of environmental surveys performed outside the FACILITY.
(4) A summary of exposures received by FACILITY personnel and visitors where such exposures are greater than 25% of that allowed or recommended.
6.7.2 SPECIAL REPORTS Special reports are used to report unplanned events as well as planned major FACILITY and administrative changes. The following special reports SHALL be forwarded to the NRC addressed in accordance with 10 CFR 50.4:
(1) There SHALL be a report not later than the following working day by telephone and confirmed in writing by facsimile or similar conveyance to the NRC Headquarters Operations Officer, to be followed by a written report within 14 days, that describes the circumstances of any of the following events:
- a. Release of radioactivity from the SITE above allowed limits. Such an occurrence SHALL be immediately reported to the Level 2 Reactor Administrator.
- b. Abnormal and significant degradation in reactor fuel, cladding, or coolant boundary, which could result in exceeding prescribed radiation limits for personnel or the environment.
(2) There SHALL be a written report within 30 days to the NRC for:
- a. Permanent changes in the FACILITY organization involving Level 1 or Level 2 management.
- b. Significant changes to the transient or accident analysis as described in the CSA 008N0128.
(3) There SHALL be a notification made to the NRC by telephone not later than 10 days following any combination of failures in equipment or administrative radiological work controls that result in a worker being assigned an unplanned dose equal to or greater than 100 mrem Total Effective Dose Equivalent (TEDE) or a spill of more than 1,000 gallons of contaminated liquid waste on uncovered (bare) soil.
In the Roadmap, section C, item 28, Section 6.7, Reports, the licensee states that the proposed changes to TS 6.7 include deletions of requirements that are not applicable in the possession-only status or relocations of requirements from current TS 6.6 to proposed TS 6.7.
The operating report was proposed to be changed to routine report and submitted every 24 months rather than annually, consistent with the possession-only status of the facility.
Current TS 6.7.1(1) and (2) were proposed to be deleted as they reflect information relevant to an operating reactor. Current TS 6.7.1(3) was proposed to be renumbered as proposed TS 6.7.1(1) and operations to be replaced with activities consistent with the possession-only status. Current TS 6.7.1(4) was proposed to be renumbered as proposed TS 6.7.1(2) and to be edited to reflect the reporting requirements in 10 CFR 50.59(d)(2). Current TS 6.7.1(5) was proposed to be deleted since effluents are no longer produced by the NTR. Current TS 6.7.1(6) and current TS 6.7.1(7) were proposed to be renumbered. TS 6.7.2(1)a was proposed to be deleted as a safety limit violation is not applicable to the possession-only status. TS 6.7.2(1)b was proposed to be renumbered as proposed TS 6.7.2(1)a, and the proposed requirement for immediate notification to the Level 2 Reactor Administrator allowed for the deletion of current TS 6.6.2, as described above. Current TS 6.7.2(1)c.v was proposed to be renumbered as proposed TS 6.7.2(1)b because TSs 6.7.2.(1)c.i through vi are not applicable to the possession-only status and were proposed to be deleted. Current TS 6.7.2.(2)b was proposed to be edited to refer to the criticality safety analysis in CSA 008N0128. Proposed TS 6.7.2(3) provides requirements for notification of unsafe conditions.
The NRC staffs review finds that the proposed changes to TS 6.7 are consistent with the possession-only status of the NTR as the requirements deleted are applicable to a reactor authorized to operate. Therefore, the NRC staff concludes that proposed TS 6.7 is acceptable.
Current TS 6.8, Records, states:
Records MAY be in the form of logs, data sheets, or other suitable forms.
The required information MAY be contained in single, or multiple records, or a combination thereof.
6.8.1 Records to be retained for a period of at least five years or for the life of the component, whichever is less:
(1) Normal reactor FACILITY operation (supporting documents such as checklists, log sheets, etc., SHALL be maintained for a period of at least one year).
(2) Principal maintenance operations.
(3) Reportable occurrences.
(4) Surveillance activities required by the Technical Specifications.
(5) Reactor FACILITY radiation and contamination surveys where required by applicable regulations.
(6) EXPERIMENTS performed with the reactor.
(7) Fuel inventories, receipts, and shipments.
(8) Approved changes in operating procedures.
(9) Records of meeting and audit reports of the review and audit groups.
6.8.2 Records of the requalification programs Records of the requalification programs SHALL be maintained in accordance with 10 CFR 55.59(c)(5).
6.8.3 Records to be Retained for the Lifetime of the Reactor FACILITY.
Note: Applicable annual reports, if they contain all the required information, MAY be used as records in this section.
(1) Gaseous and liquid radioactive effluents released to the environs.
(2) Off-SITE environmental-monitoring surveys required by the Technical Specifications.
(3) Radiation exposure for all personnel monitored.
(4) Drawings of the reactor FACILITY.
Proposed TS 6.8, Records, states:
Records MAY be in the form of logs, data sheets, or other suitable forms. The required information MAY be contained in single, or multiple records, or a combination thereof.
6.8.1 RECORDS TO BE RETAINED FOR A PERIOD OF AT LEAST FIVE YEARS OR FOR THE LIFE OF THE COMPONENT, WHICHEVER IS LESS (1) Normal reactor FACILITY operation supporting documents (such as checklists, log sheets, etc.), SHALL be maintained for a period of at least one year.
(2) Principal maintenance activities.
(3) Reportable occurrences.
(4) Surveillance activities required by the Technical Specifications.
(5) Reactor FACILITY radiation and contamination surveys where required by applicable regulations.
(6) Experiments performed with the reactor.
(7) Fuel inventories, receipts, and shipments.
(8) Approved changes in operating procedures.
(9) Records of meeting and audit reports of the review and audit groups.
6.8.2 RECORDS OF THE REQUALIFICATION PROGRAMS Records of the requalification programs SHALL be maintained in accordance with 10 CFR 55.59(c)(5).
6.8.3 RECORDS TO BE RETAINED FOR THE LIFETIME OF THE REACTOR FACILITY Note: Applicable annual reports, if they contain all the required information, MAY be used as records in this section.
(1) Gaseous and liquid radioactive effluents released to the environs.
(2) Off-SITE environmental-monitoring surveys required by the Technical Specifications.
(3) Radiation exposure for all personnel monitored.
(4) Drawings of the NTR FACILITY.
(5) Changes to the FACILITY made pursuant to 10 CFR 50.59.
In the Roadmap, section C, item 29, Section 6.8, Records, the licensee states that the proposed change of operations to activities in TS 6.8.1(2) is to align with the possession-only status of the facility. The licensee also states that the proposed change of the titles of TSs 6.8.1, 6.8.2, and 6.8.3 to be "all caps" is to set them apart as section titles. By letter dated February 23, 2024, the licensee states that it proposed to add proposed TS 6.8.3.(5) to ensure that the record retention requirements in 10 CFR 50.59(d)(3) are met.
The NRC staffs review finds that the proposed change of operations to activities in TS 6.8 is consistent with the possession-only status of the NTR. The NRC staff also finds that the proposed TS 6.8.3(5) will help to ensure that the record retention requirements in 10 CFR 50.59(d)(3) are maintained, and that the reformatted to all caps of the titles is an administrative change. Therefore, the NRC staff concludes that proposed TS 6.8 is acceptable.
3.4 Possession Only License Shutdown Configuration In the LAR, enclosures 2 and 3, the licensee provided a safety analysis of the possession-only status of the facility. The licensee evaluated several potential accident scenarios and radioactive effluent release pathways. The NRC staffs review of the POLSC involves the licensees proposed fuel storage location in the reactor and the licensees safety analysis supporting the POLSC.
Fuel Storage in the Reactor In the LAR, enclosure 2, section C, Hazards, the licensee states that all SNM has been removed from the NTR facility except for the fuel in storage in the reactor. In the Roadmap, section A, Storage of Fuel, the licensee states that the fuel must remain in storage in the reactor during the POL period because there are no other NRC-authorized storage locations at the VNC. The licensee also states that the NTR has no designated shielded storage area for spent fuel, and storage of the fuel in the reactor provides the optimum location for security and radiation protection.
In the LAR, enclosure 2, section C, the licensee indicated that modifications to the facility will be made in accordance with the POLSC to maintain the reactor subcritical. The POLSC is required by proposed TS 5.3.1, which states that the safety and control rods will remain fully inserted and restrained from any movement by ensuring that the safety rod drive belts are removed and electrically isolated, the course control rod drive chains are removed, and the control rods are electrically isolated; the MPS will consist of sizes 0.9, 0.5, and a full width poison sheet and are installed and latched in slots #1, 2, and 5, respectively, with the MPS covers installed, locked and the keys removed from the facility; and the primary coolant system is operable and the fuel loading tank is filled with water.
The NRC staff reviewed the fuel storage location in the reactor, and finds that the requirements in proposed TS 5.3.1 (evaluated and found acceptable in SE section 3.3, Proposed Changes to the Technical Specifications for the NTR) will help to ensure that the reactor remains subcritical with the control and safety rods isolated from movement, the MPS latched securely, and the primary coolant system operable to support the conductivity monitoring to help ensure that the cladding is protected from corrosion. Based the above, the NRC staff finds the proposed fuel storage in the reactor acceptable.
Release of Radioactive Material to the Environment Due to Cladding Failure In the LAR, enclosure 2, section C, the licensee states that the fuel cladding thickness has been reduced by half the original thickness due to 60 years of operation. Additional reduction or degradation of the cladding is mitigated by ensuring that the conductivity of the primary coolant water is maintained below 10 micro-siemens per centimeter (µS/cm). Proposed TS 3.3.4 states that, [t]he specific conductivity of the primary coolant water SHALL be maintained less than 5 µS/cm when averaged over the four most recent quarterly readings. Corrective action SHALL be taken if this condition is not met.
In the LAR, enclosure 2, the licensee also states that any increase in radioactivity in the primary coolant water would be detectable by the stack monitoring system, which is operated as needed, and at all times by the reactor cell area radiation monitor. The requirements for the stack monitoring system are provided in proposed TS 3.7.3, which states that the stack particulate activity monitor shall be operating when any evolution is performed in the facility that could generate an airborne concentration greater than one DAC in the reactor cell and will alarm at 2.0E-9 µCi/cc or less. If the monitor is not operable, all evolutions that could cause such airborne releases shall be discontinued within the facility and corrective action taken to restore functionality. The requirements for the reactor cell area radiation monitor are provided in proposed TS 3.7.1, which states that an operational area radiation monitor is required in the reactor cell and will alarm at 10 mr/hr or less. An operational area radiation monitor shall include: instrument readout that is visible in the control room; a gamma-sensitive instrument; a local audible alarm; and an alarm indication at a remote monitoring location.
The NRC staff reviewed the scenario of radioactive release due to fuel cladding failure and finds that the requirements in proposed TS 3.3.4 (evaluated and found acceptable in SE section 3.3) will help to ensure that the primary coolant water chemistry conductivity will continue to be monitored and maintained below 5 µS/cm to ensure that the cladding is protected from the corrosion limit of 10 µS/cm. Additionally, proposed TS 3.7.1 and TS 3.7.3, evaluated and found acceptable in SE section 3.3, will help to ensure that any potential radioactive releases will be monitored and will alarm, which will prompt corrective action. Based the above, the NRC staff concludes that the licensees analysis of the potential for a radioactive release due to fuel cladding failure accident scenario is acceptable.
Release of Radioactive Material to the Environment Due to Primary Coolant Leakage In the LAR, enclosure 2, the licensee states that any potential leakage of primary coolant into the reactor cell would be collected in the cell sump and result in a sump high level alarm when water level reaches 1 foot in the sump. This water would then be pumped to the 500-gallon holding tank as described in SAR section 5.2, Primary Coolant System. A substantial loss of coolant would cause a low-level fuel loading tank alarm at 8 inches below the tank overflow as described in SAR section 5.5, Primary Coolant Makeup Water System. The reactor cell radiation monitor would then alarm at 10 mr/hr if the water level were to drop low enough in the core tank to impact water attenuation (i.e., reduction of the radiation from the fuel). All these alarms occur in the control room and remotely.
Because cooling of the fuel is not needed to maintain the reactor shut down, the secondary cooling system will be secured, and to prevent any potential heat exchanger cross-tube leakage from escaping to the retention basins, as described in SAR section 5.3, Secondary Coolant System, the secondary inlet and outlet heat exchanger isolation valves will be closed and the drain valve for the tube side of the heat exchanger will remain open so that any cross tube leakage will drain to the reactor cell sump. In addition, routine retention basin pre-discharge sampling will continue so that any leakage across the heat exchanger outlet isolation valve would be detected prior to offsite release.
The NRC staff reviewed the scenario of release of radioactive material to the environment due to primary coolant leakage and finds that the requirements in proposed TS 3.3.3, which was evaluated and found acceptable in SE section 3.3, will help to ensure that any leakage of primary coolant is identified and corrected. Further, the NRC staff reviewed the VNC Environmental Monitoring Program as part of its review of the NTR license renewal, documented its results in its SE (ML23128A353), section 3.1.7, Environmental Monitoring, and found it acceptable. Proposed TS 3.7.2 requires air sampling and dosimetry in locations specified by the VNC Environmental Monitoring Manual, which helps to ensure that any release of radioactivity to the VNC facility is identified and not discharged offsite. Based on the above, the NRC staff concludes that the licensees analysis of the potential for a release of radioactive material to the environment due to primary coolant leakage accident scenario is acceptable.
Other Accident Scenarios In the LAR, enclosure 2, section C, the licensee evaluated the release of radioactive material to the environment by the scenarios of experimental facilities and accidental explosions, both of which the licensee states are not credible since experiments are not authorized in the POLSC and no explosive materials are stored at the facility. The licensee also evaluated the release of radioactive material to the environment due to natural phenomena, which the licensee states, as described in its CSA, would involve a seismic event, but would not result in a criticality accident, and as stated in section 13.4 of the SAR, would not result in any member of the public exceeding the limits in 10 CFR 20.1301.
The NRC staffs review finds that proposed TS 3.8, Experiments, required provides that experiments are not authorized, and finds that explosives are no longer stored at the facility, thus rendering potential accident scenarios involving experimental facilities and explosions not possible. The NRC staff reviewed the CSA (evaluated and found acceptable in SE section 3.5) and finds that the licensee has imposed requirements in proposed TS 5.3.1 that prevent an accidental criticality caused by a seismic event. The NRC staff also finds that no member of the public would exceed the limits in 10 CFR 20.1301. Based on the above, the NRC staff concludes that the licensees analysis of the potential release of radioactive material to the environment by the scenarios of experimental facilities, accidental explosions, and natural phenomena is acceptable.
3.5 Criticality Safety Analysis In the LAR, enclosure 3, the licensee details the CSA used to demonstrate that in the POLSC the facility would remain subcritical following changes in the configuration based on credible accident scenarios, which would most likely result from a seismic event.
The licensee states that the POLSC entails, in part, the NTR fuel in the reactor within the aluminum can and the safety and control rods fully inserted in the NTR core and secured by electrically isolating the drive motors and removing the drive belts, making inadvertent withdrawal of the rods impossible. There are four safety rods, two coarse control rods, and one fine control rod in the reactor. The poison section of each safety rod is 20 inches long and has a 1/2-inch diameter core of solid boron carbide within a stainless-steel tube. The poison section of the coarse control rods is 16 inches long with a 1/2-inch diameter core of solid boron carbide within a stainless-steel tube. The poison section of the fine control rod is 18 inches long and has a 0.365-inch diameter core of solid boron carbide within a stainless-steel tube. The MPS contain cadmium, which is a highly effective neutron absorber (poison), sandwiched between aluminum plates. Three MPS are also used for the POLSC, and they are restrained in place by locks attached to their covers, with the keys removed from the facility, precluding inadvertent movement of the poison sheets. The MPS are located in positions 1, 2, and 5, with thicknesses of 2.5 inches (9/10 of a sheet), 1.375 inches (1/2 of a sheet), and 2.75 inches (full width sheet),
respectively. The core tank remains full of water and the primary coolant system is operable.
The requirements to maintain the POLSC in this manner are in proposed TS 5.3.1.
In the Roadmap, section C, item 20, the licensee provided illustrations of the control and safety rods drive mechanisms to demonstrate that electrically isolating them and removing the drive belts disabled the rods and prevents any movement, thereby precluding the reactor from achieving a critical configuration while in the POLSC.
The NRC staffs review of the details provided in the licensees Roadmap, section C, item 20, Safety Rods, Course Control Rods (Two of the three control rods are course and one is fine.), and Fine Control Rod, and the accompanying figures finds that they adequately demonstrate that removing the electrical power source and drive belts leaves the drive mechanisms without any motive force, thus ensuring that the rods will not move.
CSA Scenarios and MCNP Cases As part of the CSA, the licensee evaluated all credible potential accidents from the POLSC (baseline) configuration following the completion of a Hazards and Operability Analysis (HAZOP). The HAZOP included changes to the control rods, safety rods, MPS, moderator, and geometry within the reactor. As stated in the SAR, the only credible external event for the NTR is a seismic event. As such, the CSA modeled various combinations of rods inserted, including a scenario where all rods are removed. A seismic event is also considered in the movement of the MPS. The sheets are mechanically restrained within their slots so that they will not move relative to the core during a seismic event. The licensee also evaluated flooding and the removal of graphite plugs. In all, twelve different configurations (cases) were modeled using the Los Alamos National Laboratory Monte Carlo Neutron Transport (MCNP) 6-Version 2 computer code. The 12 cases are described in the following sections of this SE.
Movable Neutron Poison Reactivity Worth Cases The licensee evaluated various configurations of removing the poison reactivity from the safety and control rods and the MPS. Case 1 is the POLSC (baseline) configuration. Cases 2 and 3 involve removal of various MPS with the safety and control rods inserted. Case 4 involves removal of all MPS. Cases 5 and 6 involve removal of the highest worth safety rod (SR-2) and removal of the highest worth coarse control rod (CR-2), respectively. Case 7 involves all rods withdrawn.
SAR section 4.2.2, Control and Safety Rods, states that the reactivity worths of the individual rods are as follows:
Control Rods Safety Rods Coarse Rod #1
$0.59 Safety Rod #1
$0.82 Coarse Rod #2
$1.01 Safety Rod #2
$1.10 Fine Rod
$0.60 Safety Rod #3
$1.07 Safety Rod #4
$0.87 SAR section 4.4.2, Reactor Core Physics Parameters, states that the reactivity parameters are based on an average effective delayed neutron fraction () of 0.00704.
Moderator Reactivity Worth Cases The licensee also evaluated flooding cases which involve filling the available MPS slots with water (case 8), filling the horizontal test facility and fuel chute plug with water (case 9), and increasing the primary coolant temperature from 75 to 124 degrees Fahrenheit (case 12). The licensee utilized MCNP to model the various cases and used equation 3-1 from the CSA.
Testing Facility and Fuel Chute Graphite Reactivity Worth Cases The licensee also evaluated cases involving removing the graphite from the vertical facility (case 10), as well as removing the horizontal facility and fuel chute plug graphite blocks (case 11).
The licensee indicates that the vertical facility is a 4-inch by 5-foot-long aluminum can that extends from the top to the bottom of the reflector within the graphite pack. The horizontal facility is a 5-inch diameter hole through the horizontal axis of the reactor. The fuel chute is a rectangular aluminum chute approximately 30 inches long, 20 inches wide, and 3 inches high.
These facilities are filled with graphite when not being used for testing.
The 12 cases described above are summarized in CSA, table 4-1, MCNP Case Descriptions, and reproduced below:
Case Number MPS Position Safety Rod Position Control Rod Position Additional Notes 1*
9/10 MPS in P1 1/2 MPS in P2 Full MPS in P5 Fully Inserted Fully Inserted 2
9/10 MPS in P1 1/2 MPS in P2 Fully Inserted Fully Inserted 3
1/2 MPS in P2 Fully Inserted Fully Inserted 4
N/A Fully Inserted Fully Inserted 5
9/10 MPS in P1 1/2 MPS in P2 Full MPS in P5 SR-2 out Fully Inserted 6
9/10 MPS in P1 1/2 MPS in P2 Full MPS in P5 Fully Inserted CR-2 out 7
9/10 MPS in P1 1/2 MPS in P2 Full MPS in P5 Fully Withdrawn Fully Withdrawn 8
N/A Fully Inserted Fully Inserted MPS space flooded 9
9/10 MPS in P1 1/2 MPS in P2 Full MPS in P5 Fully Inserted Fully Inserted Horizontal facility and fuel cute flooded Case Number MPS Position Safety Rod Position Control Rod Position Additional Notes 10 9/10 MPS in P1 1/2 MPS in P2 Full MPS in P5 Fully Inserted Fully Inserted Horizontal facility and fuel chute graphite plug removed 11 9/10 MPS in P1 1/2 MPS in P2 Full MPS in P5 Fully Withdrawn Fully Withdrawn Vertical facility graphite plug removed 12+
9/10 MPS in P1 1/2 MPS in P2 Full MPS in P5 Fully Inserted Fully Inserted Primary coolant temperature increased from 75°F to 124°F
- This is the baseline configuration
+No MCNP case was completed for this case. The k-eff is estimated based on the reactivity change from Eq. 3-1 for the baseline configuration.
The licensee provided the results of the cases in CSA, table 4-3, Reactivity Subcriticality Results, reproduced in the table below:
Case Number Keff Keff + 2 Margin to Critical ($)
1 0.92707 0.00013 0.92733
+9.82 2
0.93840 0.00011 0.93862
+8.27 3
0.94845 0.00011 0.94867
+6.88 4
0.95540 0.00012 0.95564
+5.92 5
0.93248 0.00012 0.93272
+9.08 6
0.93332 0.00012 0.93356
+8.96 7
0.96852 0.00011 0.96874
+4.12 8
0.95421 0.00011 0.95443
+6.09 9
0.88539 0.00011 0.88561
+15.16 10 0.91363 0.00012 0.91387
+11.67 11 0.96179 0.00012 0.96203
+5.05 12+
0.92757
+9.79
+No MCNP case was completed for this case. The k-eff is estimated based on the reactivity change from Eq. 3-1 for the baseline configuration.
The NRC staffs review of the CSA finds that the licensees analysis included all reasonable perturbations of the POLSC (baseline case) configuration, including the extreme configurations where all the MPS are removed (case 4), and all control and safety rods withdrawn (case 7).
The NRC staff finds that MCNP is a standard industry code and consistently used for nuclear reactor reactivity calculations, and is acceptable. The NRC staff also reviewed the benchmarking and validation of the MCNP computer code and found it acceptable for evaluating the multiplication factors of the NTR configuration. The NRC staff finds that the reactor remains subcritical in all instances. The NRC staff finds that the results provided in table 4-3 of the CSA demonstrate that there is adequate shutdown reactivity of the POLSC by several dollars. The POLSC configuration (baseline) case 1 has a margin to critical of +$9.82, which is several multiples greater than the operating limit in current TS 3.1.3, Minimum Shutdown Margin, of $1.00.
Based on the above, the NRC staff finds that the CSA demonstrated that the POLSC (baseline case), as required by proposed TS 5.3.1, is acceptable for the possession-only status of the NTR.
3.6 Proposed Certified Fuel Handler Training and Qualification Program In the LAR, section E, Staffing Changes, the licensee states that except for defueling the reactor, fuel handling operations are not permitted. Further, proposed TS 6.1.3 states that fuel movement prior to defueling is not permitted. In the LAR, the licensee requests to replace the requirements for NRC-licensed operators with requirements for CFHs in preparation for final defueling.
By letter dated February 23, 2024 (ML24054A539), the licensee provided revision 1 to the CFHTP, which replaced, in its entirety, the initial CFHTP provided by letter dated January 26, 2024.
The NRC staff previously approved a CFHTRP for the Aerotest Radiography and Research Reactor (ARRR) with the issuance of Amendment No. 6 to Facility Operating License No. R-98, dated December 6, 2021 (ML21242A463), and the basis for the NRC staffs finding of the acceptability of the use of CFHs and approval of a CFHTRP for a research reactor are detailed in section 3.7.1, Background Information and Aerotests Use of Certified Fuel Handlers, of the related safety evaluation.
Since the licensee is also requesting a POL for the NTR, where the fuel will not be moved until final defueling to support its removal from the facility and the NTR will not be authorized to operate, the NRC staff finds that there will be no need for licensed operators at the NTR.
Further, given the above information, the NRC staff finds that the underlying regulatory basis for the acceptability for an NTR CFHTP is almost identical to the regulatory basis described in the ARRR Amendment No. 6, and the staff therefore incorporates that description herein by reference.
Since there are no regulatory requirements that would specify the contents of a CFH training and qualification program, and no guidance specific to preparing and reviewing a CFHTP for a permanently shut down non-power reactor with non-licensed operators, the NRC staff is using the following regulations, guidance, and standards to inform its evaluation of the acceptability of the NTR CFHTP:
10 CFR 55.4 10 CFR 55.53, Conditions of licenses 10 CFR 55.59, Requalification NUREG-1537, Parts 1 and 2 ANSI/ANS-15.4-2016 The NRC staff also conducted its review of the NTR CFHTP based on the following broad-scope objectives, which are similar to those that have been used in evaluating the use of CFH programs for permanently shut down and defueled nuclear power reactors:
(1) Ensuring that trained individuals have requisite knowledge and experience in spent fuel handling and storage.
(2) Ensuring that trained individuals have requisite knowledge and experience in reactor decommissioning.
(3) Ensuring that trained individuals are capable of evaluating plant conditions and exercising prudent judgement for emergency action decisions.
Each NTR CFHTP section is provided below in italic font, followed by the NRC staffs evaluation of that section.
CFHTP section 1, OBJECTIVE, states:
This document defines the personnel and training requirements of the Certified Fuel Handler (CFH) program at the Nuclear Test Reactor (NTR).
The NTR no longer requires licensed reactor (RO) or senior reactor operators (SRO). By utilizing the existing core reel assembly as the fuel storage location and modifying the reactor systems to a Possession Only License Shutdown Configuration (POLSC) as defined by Technical Specification 5.3.1, the controls of the reactor can no longer be manipulated pursuant to 10 CFR 50.54(i). Additionally, in the POLSC, ample negative reactivity is permanently affixed in the core so that the removal of fuel from the reactor (defueling) poses no potential for prompt criticality and defueling of the NTR is not an alteration of the core as discussed in 10 CFR 50.54(m)(2)(iv). Since licensed operators are no longer required to manipulate the controls of the reactor or to perform core alterations, they are to be replaced by the Certified Fuel Handler (CFH) and the Certified Fuel Handler Supervisor.
The selection and training of the CFH and CFH Supervisor provides an appropriate level of oversight commensurate with the reduced risks and relative simplicity of the facility systems needed for safe storage of spent fuel, including safe defueling, handling, and storage of spent fuel, and response to plant emergencies. CFHs must follow relevant technical specifications and approved fuel handling procedures. The CFHs are obligated to know, practice, and follow, when necessary, the facility safety and security programs.
The NRC staff reviewed the proposed CFHTP section 1 and finds that it provides an appropriate description of why licensed operators are no longer required based on the new POLSC as defined by proposed TS 5.3.1. Therefore, the NRC staff concludes that CFHTP section 1 is acceptable.
CFHTP section 2, EXPERIENCE/QUALIFICATIONS, states:
2.1. CFH Supervisor (Level 3 Certified Fuel Handler Supervisor) - shall have at least 2 years of experience working in radiologically controlled environments and understand ALARA principles. Sufficient mechanical dexterity is required as evaluated by the Reactor Administrator (RA). At a minimum, requires a high school diploma or successful completion of a GED test. Maintain health/medical requirements required for the CFH job. The CFH Health Questionnaire will be used to assess health/medical requirements.
2.2. CFH (Level 4 Certified Fuel Handler) - Requires a high school diploma or successful completion of a GED test. Sufficient mechanical dexterity is required as evaluated by the RA. Maintain health/medical requirements required for the CFH job. The CFH Health Questionnaire will be used to assess health/medical requirements.
2.3. Reactor Administrator (Level 2 Reactor Administrator) - experience requirements for this role are stated in Technical Specification (TS) 6.1.4.(1).
The NRC staff reviewed the proposed CFHTP section 2. The NRC staff finds that CFHTP section 2.1 provides a description of the CFH Supervisor (Level 3) experience and qualification requirements that is reasonable given the role of the CFH Supervisor; helps ensure that the CFH Supervisor has experience and/or education that is relevant to the role; generally exceeds the guidance in ANSI/ANS-15.4-2016 for background qualifications for the Level 3 position (i.e., Reactor Supervisor) at an operating research reactor; and is consistent with proposed TS 6.1.2. In addition, the NRC staff finds that CFHTP section 2.1, in conjunction with CFHTP section 6, provides requirements that will help ensure that CFH Supervisors are medically qualified. Therefore, the NRC staff finds that CFHTP section 2.1 is acceptable.
The NRC staff finds that CFHTP section 2.2 provides a description of the CFH (Level 4) qualification requirements that is reasonable given the role of the CFH and is consistent with proposed TS 6.1.2. In addition, the NRC staff finds that CFHTP section 2.2, in conjunction with CFHTP section 6, provides requirements that will help ensure that CFHs are medically qualified.
Therefore, the NRC staff finds that CFHTP section 2.2 is acceptable.
The NRC staff finds that CFHTP section 2.3 provides a description of the experience and qualifications required for the Reactor Administrator (Level 2) as indicated in the experience requirements for this role stated in proposed TS 6.1.4(1). As indicated in CFHTP section 3.4, the Reactor Administrator may hold the concurrent position of CFH Supervisor as long as all requirements for the CFH Supervisor per CFHTP section 2.1 are maintained. Therefore, the NRC staff finds that CFHTP section 2.3 is acceptable.
Based on the above, the NRC staff concludes that CFHTP section 2 is acceptable.
CFHTP section 3, ROLES/RESPONSIBILITIES, states:
3.1 CFH Supervisor - The CFH Supervisor is a non-licensed operator who has qualified in accordance with the NTR certified fuel handler training program approved by the NRC. They will supervise other CFHs and perform CFH duties.
The CFH Supervisor is responsible for fuel handling operations and ensures the fuel handling operations are done safely, that staffing is adequate, and that CFHs have current documented training and qualifications. The CFH Supervisor has the authority to authorize a temporary deviation to a procedure involved with fuel handling but must document that deviation and report it to the Reactor Administrator by the end of the next working day. A Certified Fuel Handler Supervisor shall be in the reactor cell during fuel handling operations.
3.2. CFH - The certified fuel handler is a non-licensed operator who has qualified in accordance with the NTR certified fuel handler training program approved by the NRC.
3.3. The CFH does not make decisions on fuel-handing, decommissioning, or radiation protection. The CFH performs all necessary hands-on fuel manipulations under the direction of the CFH Supervisor and in compliance with approved fuel handling procedures.
3.3.1. The CFH only handles fuel when needed and only handles 1 fuel element at a time.
3.3.2. Fuel handling operations include only those required for defueling and fuel shipment from the facility.
3.4. Reactor Administrator - The Reactor Administrator shall ensure that the NTR is maintained according to the facility license and applicable regulations and is responsible for security and safety of the facility. The RA or his designated, certified alternate, will oversee CFH training including the assignment of generating, administering, and scoring written and operating tests. The RA may hold the concurrent position of CFH Supervisor as long as all requirements for the CFH Supervisor per section 2.1 above are maintained by the RA.
3.5. Level 1 License Holder (see TS 6.1.2.(1)) - The License Holder shall be responsible for review/approval of completed CFH Health Questionnaires.
The NRC staff reviewed the proposed CFHTP section 3. The NRC staff finds that CFHTP section 3.1 provides an appropriate description of the CFH Supervisors roles and responsibilities that are reasonable given the role of the CFH Supervisor, consistent with CFH as defined in 10 CFR 50.2 and proposed TS 6.1.2(4). Therefore, the NRC staff finds that CFHTP section 3.1 is acceptable.
The NRC staff finds that CFHTP section 3.2 provides an appropriate description of the CFHs roles and responsibilities that are consistent with the definition of the CFH in 10 CFR 50.2 and proposed TS 6.1.2(5). Therefore, the NRC staff finds that CFHTP section 3.2 is acceptable.
The NRC staff finds that CFHTP section 3.3 expands on the CFHs roles and responsibilities given in CFHTP section 3.2 and mirrors the previous NRC-approved CFH training program for a permanently shutdown and defueled ARRR non-power reactor (ML21242A463). Therefore, the NRC staff finds that CFHTP section 3.3 is acceptable.
The NRC staff finds that CFHTP section 3.4 provides an appropriate description of the Reactor Administrators roles and responsibilities that are consistent with proposed TS 6.1.2(2). The NRC staff also finds that since CFHTP section 3.4 states that the Reactor Administrator or his designated, certified alternate will oversee CFH training including the assignment of generating, administering, and scoring written and operating tests, then all the requirements for the CFH Supervisor per CFHTP section 2.1 above apply to the Reactor Administrator meaning that another CFH Supervisor will oversee the Reactor Administrator training and testing if the Reactor Administrator is a CFH Supervisor. Therefore, the NRC staff finds that CFHTP section 3.4 is acceptable.
The NRC staff finds that CFHTP section 3.5 adds to the description of Level 1 License Holder in proposed TS 6.1.2(1) by adding to the roles and responsibilities in CFHTP section 3.5 that the License Holder shall be responsible for review/approval of completed CFH Health Questionnaires. Therefore, the NRC staff finds that CFHTP section 3.5 is acceptable.
Based on the above, the NRC staff concludes that CFHTP section 3 is acceptable.
CFHTP section 4, TRAINING PROGRAM, states:
The training phase of the Certified Fuel Handler Training Program consists of lecture, and/or self-study of topics appropriate to the handling, storage, and monitoring of nuclear fuel, and includes training on CFH tasks as well as required fundamental topics.
4.1. Lectures and Self Study Topics to be covered include:
4.1.1. Design, function, and operation of systems used in handling, storage, monitoring of nuclear fuel, and auxiliary support systems.
4.1.2. Purpose and operation of the radiation monitoring systems.
4.1.3. Radiological safety principles and procedures including radiation hazards that may arise during normal and maintenance activities.
4.1.4. Conditions and limitations of facility license, including content, basis, and importance of Technical Specifications.
4.1.5. Assessment of facility condition and selection of appropriate procedures during normal, and emergency situations.
4.1.6. Fuel handling facilities and procedures.
4.1.7. Relevant NRC regulations and ALARA principles.
4.2. Job Performance Measures (JPM) CFH Training:
4.2.1. Understand annunciators; valve, pump, and breaker status indicators; and instrument readings as necessary to determine/perform appropriate remedial actions.
4.2.2. Manipulate (or simulation of) the fuel handling tool to obtain desired results during normal, and emergency conditions.
4.2.3. Understand radiation monitoring system readings, including alarm conditions, to determine appropriate actions.
4.2.4. Understand emergency conditions and remedial actions to be implemented according to the implementing emergency plan procedures for the facility.
4.3. A comprehensive final examination shall be administered at the end of the training program to provide assurance of mastery of the skills, knowledge, and abilities required for successful performance of CFH tasks. The comprehensive examination shall include a written test and an operating test. Areas examined are described in 4.1 and 4.2.
4.3.1. The written test requires a minimum score of 80 percent to pass.
4.3.2. The operating test will consist of Job Performance Measures (JPMs).
Passing criteria for an individual JPM is that the examinee successfully completes (or simulates) the assigned task in accordance with the governing procedure without missing any critical steps. Missed or incorrectly performed critical steps are the bases for JPM failure.
4.3.3. An individual who fails to pass either the written or operating test shall not perform CFH duties until he/she has completed a remedial training program and passes an appropriate retest. Only those portions of the original written or operating test that were failed need to be reexamined.
The NRC staff reviewed the proposed CFHTP section 4. Per ANSI/ANS-15.4-2016, section 5.2, a training program shall be established at each reactor facility based on the knowledge and skill required for ROs and SROs to perform their functions safely and effectively. The guidance states that training methods may be in any combination of classroom, on-the job, and self-study, and training should include both general training (e.g., nuclear and reactor technology, general operating characteristics, and radiation protection principles) and specific training (e.g., training relevant to the individual facility and its plant systems, reactor design and operation, instrumentation and controls, safety features, procedures and TSs, and applicable rules and requirements).
The regulation at 10 CFR 55.59(c)(2) describes topics for lectures during requalification programs for ROs and SROs, stating that:
The requalification program must include preplanned lectures on a regular and continuing basis throughout the license period in those areas where operator and senior operator written examinations and facility operating experience indicate that emphasis in scope and depth of coverage is needed in the following subjects:
- i. Theory and principles of operation.
ii. General and specific plant operating characteristics.
iii. Plant instrumentation and control systems.
iv. Plant protection systems.
- v. Engineered safety systems.
vi. Normal, abnormal, and emergency operating procedures.
vii. Radiation control and safety.
viii. Technical specifications.
ix. Applicable portions of title 10, chapter I, Code of Federal Regulations.
The regulation at 10 CFR 55.59(c)(3) states that the requalification program must also include on-the-job training such that operators perform appropriate plant control manipulations during the term of the operators license, demonstrate satisfactory understanding of the operation of the apparatus and mechanisms associated with the control manipulations, have appropriate knowledge of operating procedures, are cognizant of facility design, procedure, and license changes, and review the content of abnormal and emergency procedures on a regularly scheduled basis.
Although the guidance in ANSI/ANS-15.4-2016, section 5, and the requirements in 10 CFR 55.59(c)(2) and 10 CFR 55.59(c)(3) are not applicable to the NTR CFHTP, the NRC staff reviewed the information in CFHTP sections 4.1 and 4.2 in consideration of this guidance and these requirements. The NRC staff finds that CFHTP sections 4.1 and 4.2 require lectures, self-study, and JPM CFH training as part of the CFHTP, which is consistent with ANSI/ANS-15.4-2016, section 5, recommendations for initial training of ROs and SROs.
Additionally, the NRC staff finds that CFHTP sections 4.1 and 4.2 include topics and areas that are both general and facility-specific, which is consistent with ANSI/ANS-15.4-2016, 10 CFR 55.59(c)(2), and 10 CFR 55.59(c)(3) (considering the role of the CFH and the characteristics of and types of activities conducted at permanently shut down NTR). The NRC staff finds that the CFH experience and qualification requirements in CFHTP section 2, in conjunction with the training program requirements in CFHTP section 4, will help ensure that CFHs have appropriate general nuclear and radiological safety knowledge. The NRC staff notes that the CFHTP includes topics such as familiarity with emergency planning, TSs, facility design, and radiation protection. Therefore, the NRC staff finds that CFHTP sections 4.1 and 4.2 are acceptable.
The guidance in ANSI/ANS-15.4-2016, sections 5 and 6, states that initial training and requalification programs for ROs and SROs should include written, operating, and oral examinations. Additionally, 10 CFR 55.59(c)(4) requires that requalification programs for licensed ROs and SROs include written and operating tests.
Although the guidance in ANSI/ANS-15.4-2016, sections 5 and 6, and the requirements in 10 CFR 55.59(c)(4) are not applicable for the NTR CFHTP, the NRC staff reviewed the information in CFHTP section 4.3 and finds that CFHTP section 4.3 requires successful completion of written and operating tests at the completion of CFH training, before a CFH candidate can independently perform CFH duties, which is consistent with ANSI/ANS-15.4-2016 recommendations for ROs and SROs (although ANSI/ANS-15.4-2016 states that written and operating tests should be administered by the responsible authority, e.g., the NRC, the CFHTP specifies that these tests (written and operating) are conducted by the Reactor Administrator as described in CFHTP section 3.4). The NRC staff finds that the minimum 80-percent score to pass the written examination is consistent with the guidance in ANSI/ANS-15.4-2016 section 5.5, which specifies a minimum 70-percent score. The NRC staff finds that in order to pass the operating exam as stated in CFHTP section 4.3.2, all JPMs have to be successfully completed. The NRC staff also finds that the CFHTP requirement for remedial training and reexamination of CFH candidates that do not pass either the written or operating exam is consistent with ANSI/ANS-15.4-2016 section 5.5 and 10 CFR 55.59(c)(4) and that it helps determine areas where additional training is needed and helps ensure that any specific knowledge deficiencies are resolved before CFH duties are independently performed.
Therefore, the NRC staff finds that CFHTP section 4.3 is acceptable.
Based on the above, the NRC staff concludes that CFHTP section 4 is acceptable.
CFHTP section 5, ACTIVE/INACTIVE STATUS, states:
To maintain active status, each CFH shall:
5.1. Successfully complete the required training prior to, but within six months of participating in defueling the NTR.
5.2. Maintain health/medical requirements required for the CFH job. The CFH Health Questionnaire will be used to assess health/medical requirements prior to participating in defueling the NTR.
5.3. Be cognizant of any changes to any part of the requirements and obligations for safe and secure fuel handling. Changes made in procedures and the facility shall be reviewed before participating in defueling the NTR.
5.4. Participate in the annual emergency plan drill and participate in drill critiques according to the site radiological emergency plan.
The NRC staff reviewed the proposed CFHTP section 5. The regulation at 10 CFR 55.53(e) specifies requirements for ROs and SROs to maintain active status, stating that at a minimum, research and test reactor SROs and ROs are required to actively perform the functions of an operator or senior operator for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per calendar quarter. The regulation at 10 CFR 55.53(f) discusses requirements for inactive operators to return to active status, stating that at a minimum, research and test reactor SROs and ROs must complete the functions of an RO or SRO for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, under the direction of an active RO or SRO as appropriate, before returning to active status. The NRC staff finds that CFHTP section 5.1 provides an appropriate description of the CFH active/inactive status requirements that is consistent with the general intent of 10 CFR 55.53 by completing the required CFH training prior to, but within 6 months of participating in defueling the NTR. This will help ensure that CFHs properly perform their duties because proposed TSs 3.1 and 4.1 state that fuel handling in support of defueling is the only activity allowed and proposed TS 6.1.3 states that fuel movement prior to defueling is not permitted and because defueling, which involves the removal of the fuel from the facility, would not be on a timeframe that would implicate the concerns underlying the active/inactive status requirements. Therefore, the NRC staff finds that CFHTP section 5.1 is acceptable.
The NRC staff finds that CFHTP section 5.2, in conjunction with CFHTP section 6, provides requirements that will help ensure that CFHs are medically qualified and is, therefore, acceptable.
The NRC staff finds that CFHTP section 5.3 provides requirements that will help ensure that CFHs are aware of applicable changes in the facility and procedures, which is consistent with the intent of guidance in ANSI/ANS-15.4-2016, section 6.1, which states that the program shall reflect facility modifications and changes in procedures. Therefore, the NRC staff finds that CFHTP section 5.3 is acceptable.
The NRC staff finds that CFHTP section 5.4 provides requirements that will help ensure that CFHs are adequately familiar with procedures for handling emergencies and is, therefore, acceptable.
Based on the above, the NRC staff concludes that CFHTP section 5 is acceptable.
CFHTP section 6, CFH HEALTH QUESTIONNAIRE AND REVIEW, states:
6.1. All CFH applicants must fill out the NTR CFH Health Questionnaire when they apply to become a CFH.
6.2. All CFHs must fill out the NTR CFH Health Questionnaire within six months of participating in the defueling of the NTR.
6.3. The Level 1 License Holder or his designate, will review the CFH Health Questionnaire to determine that the candidate's medical condition is not such that it might cause operational errors that could endanger other plant personnel or the public health.
The NRC staff reviewed the proposed CFHTP section 6, which requires that CFH applicants complete a Health Questionnaire initially at the time of their application and that CFHs complete the questionnaire within 6 months of participating in the defueling of the NTR. The Level 1 License Holder or designate is responsible for determining whether the candidates medical condition is acceptable, using the Health Questionnaire. The NRC staff finds that the initial health review is generally consistent with the intent of the guidance in ANSI/ANS-15.4-2016, section 7, which states that medical examinations should be conducted prior to initial licensing and no less than every 2 years thereafter and that the physical condition and general health of the research reactor operator should be such that they are capable of properly carrying out their duties under normal, abnormal, and emergency conditions. The NRC staff finds that the licensee has committed to performing the medical review using the Level 1 License Holder or designate to make the determination as to whether a CFHs medical condition is acceptable.
The NRC staff finds that this is acceptable because of the limited role of the CFH as described above, because of the low risk-significance of the activities of the CFH for a POL shutdown configuration, and because NTR procedure 4.4, Reactor Fuel Unloading, section 4.2 requires a Radiation Monitor Technician, in addition to the CFH Supervisor per CFHTP section 3.1, to monitor all fuel movements. The NRC staff also finds the periodicity of at the time of application and within 6 months of participating in defueling acceptable because proposed TSs 3.1 and 4.1 state that fuel handling in support of defueling is the only activity allowed and proposed TS 6.1.3 states that fuel movement prior to defueling is not permitted and because defueling, which involves the removal of the fuel from the facility, would not be on a timeframe that would implicate more frequent reviews.
Based on the above, the NRC staff concludes that CFHTP section 6 is acceptable.
CFHTP section 7, RECORDS, states:
Records of the training certification program will be maintained to document each CFH in the program. A summary document (log) will be maintained for each CFH that includes entries to support the CFH active-duty status, attendance dates for lectures, and references for any on-the-job training activities. Records will also include copies of the written and operating tests with the answers given by each CFH. Also, any additional training given in areas where CFH exhibited deficiencies. Records will be maintained until all fuel is shipped out of the facility and CFHs/CFH Supervisor are no longer needed.
The NRC staff reviewed the proposed CFHTP section 7. The regulation at 10 CFR 55.59(c)(5) and the guidance in ANSI/ANS-15.4-2016, section 9 provides requirements and recommendations related to documentation and records for requalification programs for ROs and SROs. Although these requirements and recommendations are not applicable for the NTR CFHTP, the NRC staff reviewed the information in the CFHTP in consideration of these regulations and recommendations. The NRC staff finds that the information in CFHTP section 7 is consistent with 10 CFR 55.59(c)(5), ANSI/ANS-15.4-2016, section 9, and proposed TS 6.8.2 and that it will help ensure that the licensee documents and maintains appropriate records of its CFHTP.
Based on the above, the NRC staff concludes that CFHTP section 7 is acceptable.
Based on its review as summarized in SE section 3.6, the NRC staff finds that the NTR is not required to have NRC-licensed operators or an NRC-approved operator requalification program following the issuance of the proposed possession-only license amendment and that the licensee has proposed an acceptable approach of designating CFHs and using an NRC-approved CFHTP to ensure that it has sufficiently trained and qualified staff to safely conduct the defueling activity required by the proposed TSs. Additionally, as discussed in SE section 3.3, the NRC staff has found that the licensee has proposed an acceptable CFHTP and acceptable TSs that designate appropriate requirements related to CFHs and defueling.
Therefore, based on the above, the NRC staff concludes that the proposed elimination of NRC-licensed ROs and SROs and the use of CFHs and CFH Supervisors for defueling is acceptable.
4.0 ENVIRONMENTAL CONSIDERATION
The amendment relates, in part, to changes in recordkeeping, reporting, or administrative procedures or requirements; changes in the position and titles of officers of the licensee; and changes in the format of the license and other editorial, corrective, or minor revisions. The amendment also relates, in part, to changing requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20.
Pursuant to 10 CFR 51.22(b), no environmental assessment or environmental impact statement is required for any action within a category of actions listed in 10 CFR 51.22(c), for which the Commission has declared to be a categorical exclusion by finding that the action does not individually or cumulatively have a significant effect on the human environment.
The regulation at 10 CFR 51.22(c)(10) states, in part, that issuance of an amendment that (ii)
Changes recordkeeping, reporting, or administrative procedures or requirements; (iv)
Changes the name, position, or title of an officer of the licensee or permit holder, including but not limited to, the radiation safety officer or quality assurance manager; or (v) Changes the format of the license or permit or otherwise makes editorial, corrective or other minor revisions, including the updating of NRC approved references meets the definition of a categorical exclusion.
The regulation at 10 CFR 51.22(c)(9) states, in part, that issuance of an amendment that changes a requirement with respect to the installation or use of a facility component located within the restricted area, as defined in 10 CFR Part 20, meets the definition of a categorical exclusion, provided that the proposed change satisfies each of the following criteria:
(i)
The amendment or exemption involves no significant hazards consideration;
[10 CFR 51.22(c)(9)(i)]
Pursuant to 10 CFR 50.92, Issuance of amendment, paragraph (c), the Commission may make a determination that a license amendment involves no significant hazards consideration if operation of the facility, in accordance with the amendment, would not:
(1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or [10 CFR 50.92(c)(1)]
The proposed changes to Renewed Facility Operating License No. R-33 and the appendix A, Technical Specifications, would amend the operating license to a POL.
In the LAR, the licensee stated that the NTR would be permanently shut down on or before December 31, 2023, and proposed changes to the renewed facility operating license consistent with a POL, including the elimination of the authority to operate the reactor. By letter dated January 26, 2024, the licensee informed the NRC that it had permanently ceased operation of the NTR on December 21, 2023. In preparation for the possession-only status of the facility, the licensee removed all of the non-fuel SNM and other NRC licensed materials from the facility. The licensee left the SNM-bearing fuel assemblies in the reactor and implemented physical modifications to the control and safety rods, and manual poison sheets to ensure that the reactor would remain subcritical for the duration of the possession-only condition. Consistent with the proposed TSs, the fuel assemblies will not be moved until shipment for final removal from the VNC site (i.e., defueling).
Most postulated accident scenarios previously evaluated assume that the NTR is operating at the time of the accident initiation; therefore, these accidents are no longer applicable since reactor operation will be prohibited. However, in the LAR, the licensee states that all SNM and non-reactor related byproduct material has been removed from the facility except for the fuel in the reactor (storage). Given that the facility will be in a possession-only status with no authority to move the fuel prior to final shipment from the facility (i.e., defueling), the licensee determined that the most credible accident scenario would involve a seismic event.
In the LAR, enclosure 3, the licensee provides a CSA that modeled twelve different core configuration scenarios involving the loss of control and safety rods due to a seismic event. In accordance with proposed TS 5.3.1, the control and safety rods have been disabled and are restrained from movement in the possession-only condition. Thus, the probability of movement during a seismic event is much less than in the operating condition. Further, the results of the CSA demonstrate that the reactor core remains subcritical in all scenarios, including the loss of all rods (control and safety). Thus, the consequences of an accident from a seismic event are significantly less than before the license amendment and the possession-only condition. Based on the above, the NRC staff finds that the proposed amendment would not increase the probability or consequences of an accident previously evaluated.
(2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or [10 CFR 50.92(c)(2)]
The proposed POL does not create the possibility of a new or different kind of accident from any accident previously evaluated because the fuel will remain in storage, in the reactor, with no movement authorized until final disposition of the fuel and removal from the facility (i.e., defueling). Only a seismic event could potentially alter the position of the control and safety rods. However, the licensee has taken measures to ensure that the rods are secured from any movement. Further, the results of the licensees CSA demonstrate that if the rods were to move due to a seismic event, even with all rods withdrawn, the reactor would continue to remain subcritical.
(3) Involve a significant reduction in a margin of safety. [10 CFR 50.92(c)(3)]
The proposed POL would remove the authorization to operate the reactor and would require the licensee to maintain the fuel elements in storage, with no movement authorized, until final removal from the facility (i.e., defueling). Elimination of reactor operation significantly reduces the fission product inventory in the stored fuel elements (through radioactive decay) and thus reduces any consequences from any postulated release of radioactive material. Given that the facility will not operate and that the fuel will be maintained in storage and not moved until final disposal, the margin of safety as provided in the current TSs (for operation) is not significantly reduced (but, rather, greatly increased in the proposed possession-only TSs) as a result of the POL.
Based on the above, the NRC staff concludes that the amendment involves no significant hazards consideration.
(ii)
There is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite; and [10 CFR 51.22(c)(9)(ii)]
The NTR permanently ceased operation on December 21, 2023. Therefore, the generation of fission products also ceased and radioactive decay will lessen the potential source term available to be released offsite. Further, no experiments will be authorized in the possession-only condition, so the potential for the creation of any different types of effluents effectively ceases as well. Given that the POL will not allow operation and that no experiments will be authorized, there are no significant changes in the types or significant increases in the amounts of any effluents that may be released from the facility.
(iii)
There is no significant increase in individual or cumulative occupational radiation exposure. [10 CFR 51.22(c)(9)(iii)]
The POL will only authorize the storage of the fuel assemblies. The fuel assemblies will not be authorized to be moved until transfer to a shipping cask for final removal from the facility (i.e., defueling). The fuel assemblies fission product inventory will continue to decay and no experiments will be authorized in the POL. Thus, the POL will not cause any significant increase in individual or cumulative occupational radiation exposure.
Based on the above, the proposed changes to requirements with respect to the installation or use of facility components located within the restricted area are subject to categorial exclusion pursuant to 10 CFR 51.22(c)(9). The changes to recordkeeping, reporting, or administrative procedures or requirements; changes in the position and titles of officers of the licensee; and changes in the format of the license and otherwise editorial, corrective, or other minor revisions are subject to categorial exclusion pursuant to 10 CFR 51.22(c)(10)(ii), (iv), and (v),
respectively. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment is required to be prepared in connection with the issuance of this amendment.
5.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: G. Wertz, NRR M. Balazik, NRR L. Tran, NRR A. Miller, NRR K. Sullivan, NRR P. Torres, NRR J. Smith, NMSS Date: February 28, 2024