Text

GEH - Response to NRC Request for Additional Information
	ML24026A196
Person / Time
Site:	Vallecitos Nuclear Center
Issue date:	01/26/2024
From:	Murray S; GE-Hitachi Nuclear Energy Americas
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	M24O016
	Download: ML24026A196 (1)
	v • d • e

GE Hitachi Nuclear Energy

- HITACHI Scott P. Murray Mana ger, Facility Licensing

390 1 C astle Hayne R oad P.O. Bo x 780 January 26, 2024 W ilmington, NC 2840 2 USA M24O016 T (910) 8 19-5950 scott.murray@ge.com Via Electronic Information Exchange

ATTN : Document Control Desk U.S. Nuclear Regulatory Commission

Subject:

GEH Response to NRC Request for Additional Information

References:

1) NRC License R-33, Docket 0500073, General Electric Hitachi (GEH)

Nuclear Test Reactor (NTR)

2) GEH Vallecitos Nuclear Center (VNC) Site Physical Security Plan (PSP),

Revision 1 issued 10/11 /22

3) Letter, D. Hardesty (NRC) to C. Martinez (GEH), Issuance of Renewed Facility Operating License No. R-33 for the NTR, 6/29/23 (ML23128A348)

4) GEH/NRC Meeting to Discuss NTR Disablement and Possession Only, 8/15/23

5) GEH Application for Consent to Direct Transfers of Control of Licenses and Related Conforming License Amendments, 9/1/23 (ML23244A246)

6) GEH License Amendment Request - Permanent Cessation of the GE Nuclear Test Reactor (NTR) and Possession Only Authorization, 10/6/23 (ML23279A110)

7) Letter, G. Wertz (NRC) to C. Martinez (GEH), Acceptance of the Application for a Possession Only License Amendment, 10/18/23 (ML23291A108)

8) Letter, G. Wertz (NRC) to C. Martinez (GEH), Regulatory Audit Request for Information, 10/25/23 (ML23298A154)

9) Telecom between NRC officials, including Mr. Geoffrey Wertz and GEH representatives including Mr. Jeff Smyly held 1 / 18/24

On October 6, 2023 (Reference 6), GE Hitachi Nuclear Energy Americas, LLC (GEH) submitted a request for a License Amendment Request and Possession Only Authorization of GE H' s Vallecitos Nuclear Center (VNC) NRC license R-33 in Sunol, CA.

NRC convened a license amendment regulatory audit beginning October 19, 2023, and continu ing as necessary. Attached to this letter are several documents requested by the NRC (References 8 and 9) during that aud it.

Please contact me if you have any questions regarding this information.

I declare under penalty of perjury tha t the foregoing is true and correct.

Sinc2'1y, ft)

Scott Murray, Ma a er ~ /4 t?P\\

Facility Licensing (/

M240016 U.S. NRC January 26, 2024 Page 2 of 2

Proposed R-33 NTR Technical Specifications, Revision 8, dated January 2024 : Roadmap with the basis and justification for each Technical Specification change dated January 2024 : NTR Certified Fuel Handler Training and Requalification Program (CFHTRP) dated January 2024

Cc: G. Wertz, USNRC/NRR/DANU/UNPL SPM 24-007 Attachment 1 Proposed R-33 NTR Technical Specifications, Revision 8 Dated January 2024 GE HITACHI NUCLEAR ENERGY

NEDO 32765 Revision 8 January 2024

TECHNICAL SPECIFICATIONS FOR

THE

NUCLEAR TEST REACTOR FACILITY

LICENSE R-33

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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

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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 CAL/BRA TION:

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.

1 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 POSSESS/ON 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 NRG), 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 I 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 2

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.

3 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 spec ifications 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.

4 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 µSiem when averaged over the four most recent quarterly readings.

Corrective act ion 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 µSiem 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.

5 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 OPERA TING 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 DAG 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 1 0 mr/hr or less.

An operational area radiation monitor SHALL include:

Instrument readout that is visible in the control room.

6

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 OPERA TING when any evolution is performed in the FACILITY that could generate an airborne concentration greater than one DAG 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 DAG 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.

7 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 CAL/BRA TION 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 CAL/BRA TION biennially.

8 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

Specification

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 CAL/BRA TION 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 RAD/A TION AND ENVIRONMENTAL MONITORING SYSTEMS

Specification

4.7.1 AREA RADIATION MONITOR

Area Radiation monitor CHANNEL CHECK SHALL be performed quarterly and a CHANNEL CAL/BRA TION annually.

9 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 FACILITYLOCATION

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 Commission's regulations, is the area within the VNC SITE boundary. The restricted area, as defined in 1 O CFR Part 20 of the Commission 's Regulations, is the NTR FACILITY.

10 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.

11 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

Level 1 License Holder

Oversight Committee __________ _____ _ Level 2 Radiation Safety (audit and review function) Reactor Administrator Function

Level 3 Certified Fuel Handler Supervisor

-- Reporting Line s

Communication Lines Level 4 Certified Fuel Handler

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.

12 (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.

13 (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.

14 6.3 RAD/A TION 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

15 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.

16 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 CFR50.4:

(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

17 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).

18 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.

19 Attachment 2 Roadmap with the basis and justification for each Technical Specification change Dated January 2024 Page 1 of 24 GE Nuclear Test Reactor Roadmap of Changes to POL R-33 and Technical Specifications

Contents Document Purpose....................................................................................................................................... 2

A. Storage of Fuel.................................................................................................................................. 2

8. Changes to R-33 License................................................................................................................... 4

1. License Condition 2.8(1)............................................................................................................... 4

2. License Condition 2.8(2)............................................................................................................... 4

3. License Condition 2. 8(2)a............................................................................................................. 4

4. License Condition 2. 8(2)b............................................................................................................. 4

5. License Condition 2.8(2)c.............................................................................................................. 4

6. License Condition 2.8(3)............................................................................................................... 4

7. License Condition 2.8(3)a............................................................................................................. 4

8. License Condition 2.8(3)b............................................................................................................. 5

9. License Condition 2.8(3)c.............................................................................................................. 5

10. License Condition 2.C(l) Maximum Power Level.......................................................................... 5

11. License Condition 2.C(2) Technical Specifications........................................................................ 5

C. Changes to NTR Technical Specifications.......................................................................................... 6

1. Section 1, Introduction................................................................................................................. 6

2. Section 1.1, Definitions................................................................................................................. 6

3. Section 2.1, Safety Limits.............................................................................................................. 8

4. Section 2.2, Limiting Safety System Settings................................................................................ 9

5. Section 3, Limiting Conditions for Operation (LCO)...................................................................... 9

6. Section 3.1, Reactor Core Parameters.......................................................................................... 9

7. Section 3.2, Reactor Control and Safety Systems......................................................................... 9

8. Section 3.3, Reactor Coolant System.......................................................................................... 10

9. Section 3.5, Reactor Cell, Ventilation, and Confinement System............................................... 11

10. Section 3.8, Experiments............................................................................................................ 13

11. Section 4.0, General Surveillance Intervals................................................................................. 13

12. Section 4.1, Reactor Core Parameters........................................................................................ 13

13. Section 4.2, Reactor Control and Safety System......................................................................... 13

14. Section 4.3, Reactor Coolant System.......................................................................................... 14

15. Section 4.5, Reactor Cell Ventilation and Confinement System................................................. 14 Page 2 of 24 GE Nuclear Test Reactor Roadmap of Changes to POL R-33 and Technical Specifications

16. Section 4.7, Radiation Monitoring Systems and Effluents.......................................................... 15

17. Section 4.8, Experiments............................................................................................................ 15

18. Section 5.1, Site and Facility Description.................................................................................... 16

19. Section 5.2, Reactor Primary Coolant System............................................................................. 16

20. Section 5.3, Reactor Core and Fuel............................................................................................. 16

21. Section 5.4, Fissionable Material Storage................................................................................... 20

22. Section 6.1, Organization............................................................................................................ 20

23. Section 6.2, Review and Audit.................................................................................................... 21

24. Section 6.3, Radiation Safety...................................................................................................... 21

25. Section 6.4, Procedures.............................................................................................................. 22

26. Section 6.5, Experiments Review and Approval......................................................................... 22

27. Section 6. 6, Required Actions..................................................................................................... 22

28. Section 6. 7, Reports.................................................................................................................... 23

29. Section 6.8, Records.................................................................................................................... 24

Document Purpose This document is an accompanying document to the GE-Hitachi request to amendment license R-33 for possession-only. This "roadmap" explains and provides justification for the decision to store the NTR fuel in the reactor under the proposed possession-only license (POL) and identifies, explains, and provides justification for changes to the NTR R-33 License and Technical Specifications (TSs) to transition the facility from an operating license to the POL. An attempt has been made throughout the TSs to retain the numbering format of Revision 6 that was established according to the American National Standards Institute/ American Nuclear Society (ANSI/ANS) 15.1-2007, "The Development of Technical Specifications for Research Reactors".

A. Storage of Fuel All SNM has been removed from the NTR facility except for the fuel in storage in the reactor and that consequent to the NTR nuclear process. The reactor has been permanently shut down as of December 21, 2023. The fuel will remain in storage in the reactor according to this POL for the following reasons :

1) GEH is in the process of delivering ownership of the NTR facility over to NorthStar Group Services (NorthStar) for decommissioning according to an aggressive timeline. Transferring the core to the DOE, defueling the reactor, and transport and burial of the core cannot be completed while under license by GEH within that timeline.

Page 3 of 24 GE Nuclear Test Reactor Roadmap of Changes to POL R-33 and Technical Specifications

2) The VNC NRC site broad scope special nuclear material license (SNM-960) does not authorize the NTR core to be stored at either the hillside storage facility or building 102 hot cell or pool facilities. To authorize storage at hillside storage, SNM-960 license conditions 6, 7, and 8 would need to be amended. In addition, NRC security confirmatory order EA 144 dated 4/22 /15 would need to be amended.

3) The other three VNC NRC shutdown reactor possession only licenses (DPR-1, DR-10, and TR-1) do not authorize the NTR core to be stored at any of these facilities.

4) The VNC broad scope byproduct agreement state license CA-0017-01 does not authorize receipt or possession of special nuclear material (SNM). Agreement state licenses are not normally issued to authorize more than a critical mass of SNM.

5) The VNC NRC NTR license (R-33) currently authorizes the possession and use of the NTR core. A request was submitted to the NRC on 10/6/23 to amend the R-33 license to remove the authority to operate the NTR, authorize possession-only of the reactor and fuel, and remove operational requirements not needed for possession only status. As of December 21, 2023, the NTR has ceased operations and the reactor is being permanently disabled. NRC staff are currently reviewing the amendment request.

6) The NRC approved VNC physical security plan (PSP) will be revised shortly after the NTR facility is shutdown to incorporate 10 CFR 73.67(d) fixed site security requirements. Several of these security requirements are provided by the current facility configuration and provide greater security and control of the NTR core than other options. Contrary to a statement made in the Safety Analysis enclosed in GEH License Amendment Request - Permanent Cessation of the GE Nuclear Test Reactor (NTR) and Possession Only Authorization (ML23279Al 10),

submitted 10/6/2023, a revision to the PSP will not be made in support of the POL amendment. The aforementioned revision to incorporate 10 CFR 73.67(d) will be made pursuant to 10 CFR 50.54(p) without prior NRC review or approval.

7) Because NTR has no designated shielded storage area for spent fuel, any storage option other than the reactor would require redundant handling prior to shipping the fuel off-site.

Consideration was given to storing the fuel in a shielded transfer cask inside the reactor cell pursuant to 10 CFR 73( d)(2) ; however, this option poses additional security risk and does not maintain occupational radiation exposures ALARA when compared to storing the fuel in the reactor.

Page 4 of 24 GE Nuclear Test Reactor Roadmap of Changes to POL R-33 and Technical Specifications

B. Changes to R-33 License

1. License Condition 2.B(l)

The words (and punctuation), ", use, and operate" and "as a utilization facility" have been deleted to reflect that the reactor is no longer operational. It will not be used or operated as a utilization facility, but the remaining core fuel will simply be possessed in storage in the reactor core.

2. License Condition 2.B(2)

The words (and punctuation), "receive," and ", and use in connection with the operation of the reactor" have been deleted to reflect that ex isting SNM remains under possession at the NTR but will not be used in connection with the operation of the reactor and that no new SNM will be rece ived at the NTR. This SNM consists of the remaining fuel in the reactor and materials from reactor operations remaining in facility systems and components. The words "not separate" have been added to the end of the sentence fragment to eliminate its redundant use in each of the follow ing listed e lements.

3. License Condition 2.B(2)a The words "but not separate" have been removed to eliminate redundancy (as discussed in 2.8(2)). The word "and" has been added to the end of 2.8(2)a to clarify inclusivity of 2.8(2)b.

4. License Condition 2.B(2)b This license condition has been deleted in its entirety. Materials containing uranium-235 were used in experiments in which they underwent neutron exposure in the reactor. The reactor is no longer operable, experiments are no longer authorized, and all such materials have been removed from the facility.

5. License Condition 2.B(2)c Th is license commitment has been renumbered to 2.8(2)b to adjust for the deletion of the previous license condition. The words "but not separate" have been removed to eliminate redundancy (as discussed in 2.8(2)). The word "be " has been changed to "have been" to reflect that SNM produced by the operation of the reactor is now legacy material because the reacto r is no longe r operational.

6. License Condition 2.B(3)

The words (and punctuation), "receive," and ", and use in connection with the operation of the facility" have been deleted to reflect that ex isting byproduct material remains under possession at the NTR and will be used as necessary to support possession of the facility.

7. License Condition 2.B(3)a Th is license condition has been deleted in its entirety. According to amendment 20 of the R-33 license (issued August 18, 1992), a limit change from 200 curies to 2,000 cur ies of activated solids to be received, possessed, and used was approved in support of experiments performed at the NTR.

According to the letter requesting this change from GE to the NRC dated May 8, 1992, "These materials are taken to the NTR for neutron radiography." The reacto r is no longe r operable, expe riments are no longer authorized, and all such material has been removed from the facility.

Page 5 of 24 GE Nuclear Test Reactor Roadmap of Changes to POL R-33 and Technical Specifications

8. License Condition 2.B(3)b This license commitment has been renumbered to 2.B(3)a to adjust for the deletion of the previous license condition.

9. License Condition 2.B(3)c The words (and punctuation), "(except for byproduct material produced as allowed for experiments),"

have been deleted as no new experiments are authorized and all experiment-produced byproduct material has been removed from the NTR. The word "be" has been changed to "have been" to reflect that byproduct material produced by the operation of the reactor is now legacy mate rial because the reactor is no longer operational.

10. License Condition 2.((1) Maximum Power Level The word "not" and "at steady-state power levels not in excess of 100 kilowatts (thermal) in accordance with the limitations in the Technical Specifications" have been deleted to reflect that reacto r operation is not permitted by the POL. This license condition now simply states "The licensee is not authorized to operate the reactor at any power level."

11. License Condition 2.((2) Technical Specifications The words (and punctuation), ", as revised by Amendment No. 26" have been added to reflect the current license amendment. The word "operate" has been replaced with "maintain " to clarify that the reactor will not be operated.

Page 6 of 24 GE Nuclear Test Reactor Roadmap of Changes to POL R-33 and Technical Specifications

C. Changes to NTR Technical Specifications

1. Section 1, Introduction Verbiage is changed in the second paragraph of the introduction to reflect that the reactor w ill not be operated, but that reactor-related activity will be controlled in a way that protects the health and safety of the public, the environment, and on-site personnel. The areas addressed by these changes have also been revised to remove references to Safety Limits (SL), Limiting Safety System Settings (LSSS), and Limiting Conditions for Operation (LCO) as these are operating terms. The term Limiting Conditions for Operation (LCO) in section 3 of the TSs has been replaced with the term Limiting Conditions for Possession (LCP).

2. Section 1.1, Definitions Control Rod(s):

While the design of the control rods has not changed, their function has. The definition has been changed to reflect their specific function under the Possession-Only License (POL) of being fully and permanently inserted into the core and restrained to assist the Safety Rods and Manual Po ison Sheets in maintaining the reactor subcritical according to criticality safety analysis. In the POL Shutdown Configuration (POLSC - defined below), the control rods are fully inserted into the co re and restrained from any movement so they cannot be withdrawn.

Core Configuration:

This definition has been deleted as the sole core configuration under the POL is according to the POLSC.

Experiments:

This definition as well as reference to secured and movable experiments has been deleted as experiments are no longer authorized in the NTR (see TS 3.8). The definition of Experimental Facility/

Facilities has been retained as it describes physical facilities still remaining within the NTR facility.

Explosive Material:

This definition has been deleted as experiments are no longer performed and all explos ive mater ials have been removed from the NTR facility.

Flammable:

This definition has been deleted as experiments are no longer performed and all flammable materials have been removed from the NTR facility.

License, Licensed, or Licensee:

This definition has been changed to remove "Licensed" as it is not used in the body of the TSs.

Licensed Reactor Operator(s) / Reactor Operator(s) / Senior Reactor Operator(s):

This definition has been deleted as a 10 CFR 55 license is not required in a non-operating reactor. Fuel movement (defueling) will be done by Certified Fuel Handlers.

Manual Poison Sheet(s) (MPS):

The MPS were used to compensate for burnout over the operating life of the core; however, they will now be used in the Possession-Only License (POL) to ensure the reactor remains subcritical. The words Page 7 of 24 GE Nuclear Test Reactor Roadmap of Changes to POL R-33 and Technical Specifications

"ma intain adequate negative reactivity in the reactor to prevent attainment of criticality" have been added and the reference to fuel burnout removed. The POLSC is now considered a design feature and details, including numbers and positions of MPS, are prov ided in Section 5 of the TSs.

Measured Value:

This definition has been deleted as the term was only applicable to parameters having to do with reactor safety systems; specifically, TS surveillance 4.1.1, Potential Excess Reactivity, and TS 2.2, Linear Power.

These have been deleted as the reactor is no longer operational.

Possession Only License Shutdown Configuration (POLSC):

A definition for POLSC has been added to the proposed TSs to describe the facility modifications that maintain the reactor shutdown configuration. Details are included as design features in section 5 of the proposed TSs. The POLSC includes modifications that are verifiable through surveillance, including verification that manual poison sheets (MPS) remain locked in place and fully inserted control and safety rod are restrained by removing their drive linkages and e lectrically isolating their motors.

Potential Excess Reactivity:

This definition has been deleted as it is dependent on manipulation of the Control Rods - which are not operable.

Protective Actions:

This definition has been deleted as the reactor is no longer operable unde r the POLSC.

Reactivity Worth (Experiment):

This definition has been deleted as Experiments are no longer performed in the NTR and the reactor is under the POLSC.

Reactor Operating or Reactor Operation(s):

This definition has been deleted as the reactor is no longer operational under the POLSC.

Reactor Thermal Power:

This definition has been deleted as the reactor is no longer ope rational unde r the POLSC.

Reactor Safety System(s):

This definition has been deleted as the reactor is no longer operational and these systems no longer perform a function under the POLSC.

Reactor Secured:

This definition has been deleted as the reactor is permanently secured according to the POLSC.

Reactor Shutdown:

This definition has been deleted as the reactor is permanently shut down according to the POLSC.

Reactor Shutdown Configuration:

This definition has been modified to describe the singular core configuration (rods fully inserted and MPS locked in place) - which is concerned with holding the initial conditions for the bounding criticality analysis invariable by restraining movable poisons in the core with sufficient negative reactivity to keep Page 8 of 24 GE Nuclear Test Reactor Roadmap of Changes to POL R-33 and Technical Specifications

the reactor subcritical. Reactor Shutdown Configuration (RSC) differs from POLSC in that the POLSC is concerned with the modifications necessary to maintain the RSC.

Readily Available Senior Reactor Operator:

This definition has been deleted as Licensed Operators will no longer be used at the NTR under the POLSC.

Reference Core Condition:

This definition has been deleted as its only relevance was pertaining to shutting down the reactor. The reactor is now permanently shut down under the POLSC.

Safety Rods:

While the design of the safety rods has not changed, their function has. The definition has been changed to reflect their specific function under the Possession-Only License (POL) of being permanent inserted and restrained in the core to assist the Control Rods and Manual Poison Sheets in maintaining the reactor subcritical according to criticality safety analysis. In the POLSC, the safety rods are fully inserted into the core and mechanically and electrically disabled so they cannot be withdrawn.

Scram Time:

This definition has been deleted as have TS 3.2.5, and surveillance 4.2.5 since a reactor scram is not relevant to a permanently shut down reactor.

Shutdown Margin:

This definition has been deleted as the reactor is permanently shut down according to the POLSC.

Surveillance Intervals:

Quinquennial, monthly, weekly, and prior to SU have been deleted as they are no longer used in the TS surveillance interval.

True Value:

This definition has been deleted as it was only relevant to TS 2.1, Reactor Thermal Power, which has been deleted.

Unsafe Condition:

This definition has been deleted as it was connected to the reporting conditions of TS 6.1.1, which has been deleted. An additional reporting condition has been added to Section 6 (proposed TS 6.7.2(3)) to address safety concerns and aligns with criteria previously established at the VNC for existing shutdown reactors.

Unscheduled Shutdown(s):

This definition has been deleted as the reactor is permanently shutdown according to the POLSC.

3. Section 2.1, Safety Limits The applicability heading has been deleted and the introductory statement for this specification has been edited to reflect that safety limits are not applicable for a permanently shut down reactor in a POLSC. The Objective, Specification, and Basis sections of the specification have been deleted as the specification is no longer applicable. Rather than simply deleting nonapplicable specifications, this Page 9 of 24 GE Nuclear Test Reactor Roadmap of Changes to POL R-33 and Technical Specifications

format has been adopted throughout these TSs to maintain continuity in formatting according to the ANSI 15.1.

4. Section 2.2, Limiting Safety System Settings The applicability heading has been deleted and the statement for this specification has been edited to reflect that reactor safety limits are not applicable for a permanently shut down reactor in a POLSC. The Objective, Specification, and Basis sections of the specification have been deleted as the specification is no longer applicable.

5. Section 3, Limiting Conditions for Operation (LCO)

This specification is retitled as Limiting Conditions for POLSC (LCP) to reflect that the reactor has permanently ceased operation.

6. Section 3.1, Reactor Core Parameters The applicability heading has been deleted and the introductory statement for this specification has been edited to reflect that safety limits are not applicable for a permanently shut down reactor in a POLSC and a clarifying statement has been included (relative to the core) that fuel handling in support of defueling is the only activity allowed. The Objective, Specification, and Basis sections of the specification have been deleted as the specification is no longer applicable.

7. Section 3.2, Reactor Control and Safety Systems Specification 3.2.0, General, has been deleted as the need to perform a reactor shutdown is no longer applicable for a permanently shut down reactor in a POLSC.

The Applicability and Objective have been deleted as the only acceptable positions for the rods and MPS are fully inserted and restrained from any movement according to the POLSC.

Specification 3.2.1, Rods Operable, has been renamed to Rods Inoperable and edited to ensure the Reactor Shutdown Configuration is maintained by applying the conditions of the POL SC (TS 5.3.1)

- that is, fully inserted and restrained from any movement.

Specification 3.2.2, Safety Rod Withdrawal has been deleted as it is not applicable to a permanently shut down reactor and replaced with proposed specification 3.2.2, Manual Poison Sheets Secured, to reflect the repurposing of the specification from ensuring the reactor remains subcritical during startup safety rod withdrawal to maintaining the Reactor Shutdown Configuration by means of the verifiable POLSC (TS 5.3.1).

Specifications 3.2.3, Safety Rod Withdrawal Rate; 3.2.4, Control Rod Withdrawal Rate; 3.2.5, Scram Time ; and 3.2.6, Reactor Safety System and Safety Related Items, have been deleted as they are not applicable for a permanently shut down reactor in a POLSC.

Tables 3-1, Reactor Safety System - Scram, 3-2, Reactor Safety-Related Items, their bases, as well as surveillance Tables 4-1, Surveillance Requirements of Reactor Safety System Scram Instruments, and 4-2, Surveillance Requirements of Reactor Safety-Related Items (Information Instruments), have been deleted as only a few of the listed systems are being maintained to support proposed POL TSs.

These systems have been relocated to their relevant specifications and supporting surveillances. For Page 10 of 24 GE Nuclear Test Reactor Roadmap of Changes to POL R-33 and Technical Specifications

instance, Coolant Flow is addressed in proposed TS 3.3.1, Forced Flow Cooling; Reactor Cell Pressure is addressed in proposed TS 3.5.1, Reactor Cell Negative Pressure; Fuel Loading Tank Water Level is addressed in proposed TS 3.3.2, Fuel Loading Tank Full and in proposed TS 3.3.3, Fuel Loading Tank Level Alarm; and Stack Radioactivity is addressed in proposed TS 3.7.3, Stack Monitor Operability.

The Basis section for TS 3.2 has been edited to clarify that maintaining the facility according to proposed LCPs 3.2.1 and 3.2.2 ensures the reactor remains safely subcritical with adequate negative reactivity present to ensure reactor criticality does not occur.

8. Section 3.3, Reactor Coolant System This section has been retitled from Reactor Coolant System to Primary Coolant System to better align with terminology in the SAR. The Secondary Coolant System is given in Section 5.3 of the SAR as a subsystem of the Reactor Coolant System. However, since the Secondary Coolant System will not be maintained operable, a more precise title is needed for TS Section 3.3. The proposed LCP 3.3. title, "Primary Coolant System", includes the Primary Coolant System described in SAR 5.2 as well as the Primary Coolant Cleanup and Primary Coolant Makeup Water Systems described in SAR 5.4 and S.S. All of these are necessary to support the assumptions and recommendations of the criticality safety analysis. The Applicability and Objective statements have been deleted. Forced Flow Cooling (TS 3.3) is no longer necessary to provide core cooling for the operating reactor, but for corrosion/ conductivity control and monitoring pursuant to NUREG-1537, 17.2.1.2.

Specification 3.3.1, Forced Flow Cooling, has been edited to remove the cooling system function for an operating reactor above 0.1 kW as the reactor is no longer operable. The specification is repurposed to ensure that coolant flow is simply operable according to the POLSC and now states that forced primary coolant flow shall be operable. The words "light water" have been removed as they unnecessarily describe what is implicit by reactor design. The word "primary" has been added to clarify that the specification is not applicable to secondary cooling. A requirement to restore operability within 90 days has been added. 90 days has been chosen as a reasonable time to restore flow in the primary coolant system. Coolant flow is not critical to maintaining the stored reactor fuel subcritical but only supports chemical mixing and proper operation of the conductivity meter. The impact of losing flow to system corrosion control would be gradual; however, flow is also necessary for operation of the installed conductivity meter that supports proposed TS 3.3.4.

Specification 3.3.2, Core Tank Full, has been retitled to Fuel Loading Tank Full. The Fuel Loading Tank sits higher in the primary coolant system than the Core Tank and is vented to the atmosphere. Therefore, by design, the Core Tank is full if the Fuel Loading Tank is full. Because proposed LCP 3.3.3 is based on the Fuel Loading Tank alarm, the title Fuel Loading Tank Full is appropriate for proposed LCP 3.3.2. Proposed LCP 3.3.2 has been further repurposed, from ensuring adequate cooling for the reactor core to simply ensuring the system is full according to the POLSC. This is the most important specification in section 3.3 because maintaining water in the system is an assumptive condition for the criticality safety analysis.

The level need only be maintained above the low-level alarm to ensure the alarm will be operable and that remote notification will be given at the central alarm station that remedial action should be taken.

Page 11 of 24 GE Nuclear Test Reactor Roadmap of Changes to POL R-33 and Technical Specifications

Specification 3.3.3, Primary Coolant Conductivity, has been renumbered to proposed LCP 3.3.4. The LCP retains the 5 µS/cm limit but changes the reading frequency to quarterly from monthly because the less harsh (relative to reactor operations) conditions of the POLSC do not warrant verifying operability of the conductivity probe as frequently. Because system flow is necessary for operation of the conductivity probe, proposed LCP 3.3.1 requires system flow operability restoration within 90 days of failure to ensure a quarterly conductivity reading with a low probability of missing a reading as the result of a reactor coolant system failure. Since a quarterly reading is the least frequent reading that can be assured, the 5 µS/cm limit of proposed LCP 3.3.4 will be interpreted as the average of the four most recent readings. This will provide ample time to take corrective actions should the 5 µS/cm limit be challenged.

Proposed specification 3.3.3, Fuel Loading Tank Level Alarm, has been added to complement proposed LCP 3.3.2 by ensuring an alarm is maintained that provides early detection of a loss of coolant level.

Specification 3.3.3 clarifies that a local visible alarm and a remote alarm are set to actuate when water level drops to 3 feet below the overflow or higher. "Or higher" allows for setting the alarm more conservatively if deemed appropriate.

The Basis for this section has been rewritten to address the repurposing of the proposed LCPs in this section, from providing cooling of the operating reactor to prov iding for ongoing chemistry control and establishing the assumptive conditions for maintaining the reactor subcritical according to the POLSC.

The title of TS 3.3, "Primary Coolant System", includes the Primary Coolant System described in SAR 5.2 as well as the Primary Coolant Cleanup System and Primary Coolant Makeup Water System described in SAR 5.4 and 5.5. All of these are necessarily maintained to support assumptions and recommendations of the criticality safety analysis.

9. Section 3.5, Reactor Cell, Ventilation, and Confinement System The Applicability and Objective sections of this specification have been deleted as the specification is no longer applicable in the POLSC.

Specification 3.5.1, Reactor Cell Negative Pressure, has been edited to reflect that the reactor is no longer operational, and that the ventilation system now serves the primary function of minimiz ing airborne contaminants in the facility below one Derived Air Constant (DAC) of airborne radioactivity (10 CFR 20, Appendix B, Table 1, Column 3) by providing a filtered pathway to evacuate that contamination to the outside environment. An evolution in the reactor cell capable of producing one DAC or greater would require that the control room door be open for access. Therefore, establishing proper operation of the ventilation system must be done prior to commencing such an evolution by temporarily closing the control room door and verifying a reactor cell to control room negative differential p ressure of not less than 0.5 in. of water. Proper operation of the ventilation system can likewise be verified at any time by cycling the door. As evidenced by the bounding analysis provided in Chapter 13 of the SAR and criticality safety analysis performed in support of this POL, the existing ventilation system will ensure that emissions to the environment will not exceed the limits of 10 CFR 20.1301 for members of the public. Additionally, verbiage has been added to say that the "ventilation system shall be operating."

Although this is implicit in order to maintain a negative pressure in the reactor cell, including this clarifying statement allows for the deletion of the somewhat redundant LCO (LCP) 3.5.2.

Page 12 of 24 GE Nuclear Test Reactor Roadmap of Changes to POL R-33 and Technical Specifications

Specification 3.5.2, Reactor Cell Activity Release, has been deleted. It is so stated in proposed LCP 3.5.1 that the ventilation system shall be operating during performance of activities that could result in an airborne concentration of one DAC or greater in the reactor cell. This makes specification 3.5.2 redundant.

The Basis for section 3.5 has been updated to reflect that operation of the ventilation system is no longer tied to operation of the reactor, that LCO 3.5.1 is now proposed LCP 3.5.1, and to clarify that qualifying activities requiring operation of the ventilation system are those that are anticipated to generate an airborne concentration greater than one DAC in the reactor cell.

Section 3.7, Radiation Monitoring Systems and Effluents This section has been retitled to Radiation and Environmental Monitoring Systems, to reflect that the facility is no longer emitting radioactive effluents from reactor operation. This is due to several reasons addressed in the POL Safety Analysis document, including: reduced operation of the ventilation system, discontinuance of the formation of activation and fission products by the operating reactor, and limiting of potential airborne-generating activities performed under the POL.

The Applicability and Objective sections of this section have been deleted. The installed area radiation monitors, except for the reactor cell monitor, will be removed from service, and the ventilation stack gaseous activity monitor will be removed from service as discussed in the POL Safety Analysis document. Verbiage clarifying that the function of the area monitors was related to radiation safety and not reactor (nuclear) safety is also no longer necessary as the reactor is permanently shut down and it is therefore understood that the remaining area monitor provides for radiation safety.

Specification 3. 7.1, Monitoring Systems During Reactor Operations, has been deleted as it was only applicable to an operating reactor.

Specification 3.7.2, Monitoring Systems During Reactor Cell Maintenance, has been retained, but is retitled "Area Radiation Monitor" (referencing the reactor cell radiation monitor), renumbered as 3.7.1, and adds the 10 mr/hr or less alarm setpoint discussed in the POL Safety Analysis document.

"Or less" allows for reducing the setpoint as decaying background radiation allows. The word "operational" has replaced "functional" to align with terminology in NUREG-1537, Part 1, 3.7.1.

TS 3.7.3, Effluents - Environmental Monitoring, has been renamed, Environmental Monitoring, and renumbered to 3.7.2. The reference to thermoluminescent (TLD) badges has been updated to reflect the equivalent use of optically stimulated luminescent (OSL) dosimeters. Environmental monitoring for the VNC continues to be implemented under the sitewide effluent monitoring program according to the VNC Environmental Monitoring Manual (EMM).

TS 3.7.4, Effluents - Stack Release Activity, has been deleted along with Table 3-3, Stack Release Action Levels (and footnotes), as action levels based on operational effluents (activation / fission products) are no longer applicable.

TS 3.7.5, Effluents - Stack Monitoring Operability, has been renumbered to proposed LCP 3.7.3, retitled to Stack Monitor Operability, and revised to require stack monitoring only for particulate Page 13 of 24 GE Nuclear Test Reactor Roadmap of Changes to POL R-33 and Technical Specifications

activity and for evolutions identified in proposed LCP 3.5.1. Proposed LCP 3. 7.3 also includes a particulate channel alarm setpoint of 2.0E-9 µCi /cc or less. A setpoint of 5.0E-10 µCi /cc or less, was proposed in the POL Safety Analysis document; however, since the reactor was permanently shut down on December 21, 2023, dose rates have not decayed as rapidly as anticipated and it has become apparent that 5.0E-10 µCi /cc is not a realistic initial setpoint under the POL. 2.0E-9 µCi /cc is slightly above current background and can be adjusted conservatively as radioactive decay reduces background radiation levels. The words "operational" and "operability" have replaced "functional" and "functionality" to align with terminology in NUREG-1537, Part 1, 3.7.1.

The Basis for section 3.7 has been edited to replace "operations personnel" with "facility personnel" to reflect that there is no longer a reactor operations department. The Basis also describes the repurposing of the ventilation system and stack monitoring system under the POL (which is addressed in proposed TS section 3.5), which are no longer required to monitor effluence from reactor operations. The Basis also removes reference to the area radiation monitors that are no longer being maintained operable.

10. Section 3.8, Experiments The Applicability section title has been deleted and the applicability statement modified into a section introductory statement that states that experiments are no longer authorized due to the reactor being in a POLSC and that explosives are no longer stored in the facility.

The balance of section 3.8, and its Basis have been deleted as they are no longer applicable.

11. Section 4.0, General Surveillance Intervals This section has been deleted in its entirety as it is applicable to an operating reactor.

12. Section 4. 1, Reactor Core Parameters This surveillance is being maintained for consistency in formatting to mirror proposed LCP 3.1, although performance of the surveillance has been voided. The Applicability Statement title for this surveillance has been deleted and re-written as an introductory statement to reflect that section 4.1 is no longer applicable under a POLSC and that fuel handling in support of defueling is the only activity allowed relative to the core.

The balance of section 4.1 including its Basis has been deleted as the entirety of section 4.1 is not applicable to a permanently shut down reactor.

13. Section 4.2, Reactor Control and Safety System The applicability and objective statements for this specification have been deleted as the POLSC is the only mode of operation and the only objective for surveillances performed in support of proposed LCP 3.2 are those that verify the POLSC.

Specification Surveillance 4.2.1, Rods Operable, has been retitled to Rods Inoperable and establishes a semi-annual verification that both safety rods and control rods meet the conditions of the POLSC (as per definition and proposed TS 5.3.1). The existing operability test for rod movement has been deleted as the rods will remain fully inserted and restrained from movement under the POLSC.

Page 14 of 24 GE Nuclear Test Reactor Roadmap of Changes to POL R-33 and Technical Specifications

Specification Surveillance 4.2.2, Safety Rod Withdrawal, has been deleted, as rods are fully inserted and restrained from any movement the POLSC. In its place, proposed specification surveillance 4.2.2, Manual Poison Sheets Secured, has been added that requires a semi-annual verification that the MPS are locked in place and keys removed from the facility according to the POLSC.

Specification Surveillances 4.2.3, Safety Rod Withdrawal Rate; 4.2.4, Control Rod Withdrawal Rate; 4.2.5, Scram Time; and 4.2.6, Reactor Safety System and Safety-Related Items, have been deleted as they are no longer applicable under a POLSC. As previously discussed, surveillance Tables 4-1, Surveillance Requirements of Reactor Safety System Scram Instruments, and 4-2, Surveillance Requirements of Reactor Safety-Related Items (Information Instruments) have been deleted.

The Basis for section 4.2 has been revised to remove reference to deleted surveillances and to support proposed surveillances 4.2.1 and 4.2.2; that is, to verify that rods and MPS remain in the POLSC.

14. Section 4.3, Reactor Coolant System Section 4.3, Reactor Coolant System is retitled to Primary Coolant System to align with terminology used in proposed sections 3.3 and 5.2. This specification surveillance previously pointed to requirements in Tables 4-1 and 4-2. Since these Tables have been deleted, and because the primary side of the reactor coolant system will be maintained operable under the POLSC, appropriate surveillances have been proposed for this section.

There are no Applicability or Objective sections for this specification as the only mode of applicability is the POLSC.

Proposed Specification Surveillance 4.3.1, Forced Flow Cooling, has been added in support of LCP 3.3.1 to perform quarterly channel checks and annual calibrations on the primary coolant flow instrument.

Proposed Specification Surveillance 4.3.2, Fuel Loading Tank Full, has been added in support of LCP 3.3.2 to perform quarterly visual verification that the fuel loading tank is full.

Proposed Specification Surveillance 4.3.3, Fuel Loading Tank Level Alarm, has been added in support of LCP 3.3.3 to perform quarterly channel tests for the fuel loading tank level alarm.

Proposed Specification Surveillance 4.3.4, Primary Coolant Conductivity, has been added in support of LCP 3.3.4 to perform quarterly channel checks and biennial calibrations of the primary coolant conductivity instrument.

A Basis for Section 4.3 has been added to reflect the repurposing of the primary coolant system from cooling the operational reactor to verifying the POLSC and the objectives of proposed LCP section 3.3.

15. Section 4.5, Reactor Cell Ventilation and Confinement System The Applicability and Objective sections for this specification have been deleted as the only mode of applicability is the POLSC.

Specification Surveillance 4.5.1, Reactor Cell Negative Pressure, has been revised to remove its association with deleted Table 4-2 and establish a daily channel check (when the ventilation system is operating) and an annual calibration of the reactor cell differential pressure instrument.

Page 15 of 24 GE Nuclear Test Reactor Roadmap of Changes to POL R-33 and Technical Specifications

Specification Surveillance 4.5.2, Reactor Cell Activity Release, has been deleted to mirror the deletion of LCO 3.5.2 and because it is redundant to the daily channel check of proposed specification surveillance 4.7.3.

The Basis for Section 4.5 has been revised slightly. The word "Operation" was deleted at the beginning of the paragraph and replaced with "Maintaining the facility" to reflect that the reactor is not operable.

Reference to deleted TS surveillance 4.5.2 has been deleted.

16. Section 4. 7, Radiation Monitoring Systems and Effluents This section has been retitled to Radiation and Environmental Monitoring Systems, to reflect that the facility is no longer emitting radioactive effluents related to operation of the reactor.

The Applicability and Objective have been deleted as the only remaining mode of operation is the POLSC.

Specification Surveillance 4.7.1, Monitoring Systems During Reactor Operation, has been retitled, Area Radiation Monitor. References to all area radiation monitors except for the reactor cell monitor have been deleted so that the surveillance now only requires a quarterly channel check and an annual calibration for the reactor cell area radiation monitor.

Specification Surveillance 4.7.2, Monitoring Systems During Reactor Cell Maintenance, has been deleted.

LCO 3.7.1 was deleted, resulting in the renumbering of LCP 3.7.2 as LCP 3.7.1. Proposed Specification Surveillance 4.7.1 is now the appropriate surveillance for proposed LCP 3.7.1.

Specification Surveillance 4.7.3, Effluents - Environmental Monitoring, has been retitled to Environmental Monitoring ("Effluents" has been deleted.), to reflect discontinuance of monitoring of nonexistent effluent activation and fission products and has been renumbered to 4.7.2 to correspond to proposed LCP 3.7.2. Subparagraph "a" has been updated to reflect the equivalent use of OSL dosimeters or TLDs for monitoring site dose.

Specification Surveillance 4.7.4, Effluents - Stack Release Activity, has been deleted consequent to the deletion ofTS 3.7.4.

Specification Surveillance 4.7.5, Effluents - Stack Monitor Operability, has been retitled to Stack Monitor Operability ("Effluents" has been deleted.) to reflect the title of proposed LCP 3. 7.3. Since LCO surveillances 4.7.3 and 4.7.4 have been deleted, 4.7.5 has been renumbered to 4.7.3, correspondent to proposed LCP 3.7.3. Proposed surveillance 4.7.3 establishes a daily channel check and an annual calibration for the stack particulate activity monitor channel.

Basis -A Basis has been added to 4. 7 to amplify information for the extensive changes made to this section.

17. Section 4.8, Experiments This section has been retained for continuity in formatting and edited to mirror proposed LCP section 3.8. It reflects that these surveillances are not applicable according to the POLSC and that experiments, and storage of explosives are no longer authorized at the NTR.

Page 16 of 24 GE Nuclear Test Reactor Roadmap of Changes to POL R-33 and Technical Specifications

18. Section 5.1, Site and Facility Description Specification 5.1.3 has been deleted. As discussed in the Bas is, stack he ight was determined to be of sufficient height to disperse effluent exhaust upward. As per the POL Safety Analysis, effluence from activation and fission products is no longer generated at the NTR. Adequate dispersion remains important relative to proposed LCP 3.5.1; however, this is now cons idered a function of site radiation safety and not a design feature of the NTR.

19. Section 5.2, Reactor Primary Coolant System The title of section 5.2 is changed to remove the word "reactor." "Primary Coolant System" aligns with the terminology used in proposed LCP sections 3.3 and 4.3.

Specification 5.2.1, Primary System Pressure, is retitled to Primary Coolant System Pressure to align with the terminology in proposed LCP sections 3.3 and 5.2.

20. Section 5.3, Reactor Core and Fuel Specification 5.3.1, Control System, has been retitled to, Possession Only License Shutdown Configuration (POLSC). The existing content of specification 5.3.1 is irrelevant to the POLSC as safety rods and control rods will be remained fully inserted and restrained both electrically and mechanically.

Electrical isolation of the rods is described in Enclosure 3 of letter GEH License Amendment Request -

Permanent Cessation of the GE Nuclear Test Reactor (NTR) and Possession Only Authorization, submitted on October 6, 2023 (Agencywide Documents Access and Management System [ADAMS] accession no.

ML23279A110). Mechanical isolation of safety and control rods is described in "Rods Fully Inserted and Restrained" (below).

In addition to the rods being fully inserted and restrained, installed MPS will remain restrained in the reactor under the POL through defueling of the reactor. The sizes of each MPS sheet and pos ition in the reactor are detailed in this specification.

According to the POLSC, it is necessary that the Fuel Load ing tank is maintained full of water and the Primary Coolant System maintained operable. These conditions are ensured by proposed LCPs in TS section 3.3, Primary Coolant System.

Operability of the primary coolant system as required by proposed specification 5.3.1 deviates from the description in section 5. 2.2 of the SAR but is implicitly described by proposed LCPs 3.3.1, 3.3.2, 3.3.3, and 3.3.4. The primary coolant system is operable when it is full of water with an operable low-level alarm and forced flow can be established so that system conductivity can be verified to be less than 5 µS/cm.

Because of the design of the NTR, the presence of components in this configuration will not interfere with or in any way impede defueling of the reactor. The content of specification 5.3.1 has therefore been replaced with the assumed initiating cond itions for the POL criticality safety analysis. POLSC is defined in the definitions section of these proposed TSs and these conditions are inculcated in proposed LCPs in section 3 and verified maintained by supporting surve illances in section 4 of the TSs.

Page 17 of 24 GE Nuclear Test Reactor Roadmap of Changes to POL R-33 and Technical Specifications

Rods Fully Inserted and Restrained

Safety Rods:

Safety rods have a drive mechanism that electromagnetically couples (See 11 & 26 in Figure 1.) to the poison rod (1-3) and withdraws it from the core against spring (7-9) pressure. The motor (14) is attached (17-18) to the drive screw (13) through gears (16) driven by nylon belts (15). Upon a reactor scram, the springs return the poison rod to the fully inserted core position and the drive mechanism (12, 23, 24) follows and recouples to the poison rod. The drive mechanism, which can only be moved by the drive motor, physically blocks the poison rod from moving in the outward direction from the core. If the drive mechanism cannot move because its motor is electrically isolated and its nylon drive belt is removed, the rods cannot move in the outward direction.

.T.R.

S FETY ROD AD MECHANI S ROD DRIVE

Figure 1 Page 18 of 24 GE Nuclear Test Reactor Roadmap of Changes to POL R-33 and Technical Specifications

Course Control Rods (Two of the three control rods are "course" and one is fine."):

Course control rods (See 1-3 in Figure 2.) are mechanically linked (5-8) to a lead screw (10). The lead screw must turn to allow control rods to move in an outward direction from the core. The lead screw is linked (11) to a drive motor (17) through a chain gear (14-16). If the lead screw cannot turn because the motor is electrically isolated and the chain drive is removed, then the rods cannot move in the outward direction.

----, ~, _-,,_

l. Eight1/2" x 2 " Boron Ca.rblde Cylinders, Total 16 " or Polson 14. Sprocket ohaln I 5. Drive Sprocket 2, End Plugs 16. Gear Box 57 RP~ o.itput

3. Stainless Steel Cyllnder 17, Electric Motor -** hp Capacitor start

4. Extension 18, Yoke Support and Gulde Rod

5. Pinned Coupling 19. Rod - In Switch 6, Rod Stop Armature and Stator Unit {Not Used) 20. Rod - out SWltch

7. IJn tversal Jolnt 21. Gulde Rod Tube 8, 'l!lke 22, Selsyn Transmitter 9, Rod Follower 23. Selsyn Rack and Drlvs Gears 10, Lead Screw 24. SUppot1 Bracket 11, sprocket and NUl Aeeeml>Jy 2 5. Reactor North Face 12, Lead Screw Gulde Tube 13, Adjustable Stop Bolt N.TR.

COARSE CONTROL ROD AND ROD DRIVE MECHANISMS

Figure 2 Page 19 of 24 GE Nuclear Test Reactor Roadmap of Changes to POL R-33 and Technical Specifications

Fine Control Rod:

The fine control rod (See 1-3 in Figure 3.) has an integrated direct drive (containing the motor and gearing) (12-15) that turns a lead screw (10) that is mechanically coupled (5-9 & 11) to the associated poison rod. If the lead screw cannot turn because its motor is electrically isolated, then the rod cannot move in the outward direction.

OL s s

Figure 3

Specification 5.3.2, Reactor Fuel, has been deleted as it provides design criterion for future refueling -

which will not occur under the POL.

Specification 5.3.3, Core Reel Assembly, has been edited to delete the qualifying statement that the reactor be shutdown, which is evident under the POL. A new qualifying statement has been added to allow rotating of the core reel assembly only during authorized fuel handling activities. This specification is renumbered as proposed specification 5.3.2.

Specification 5.3.4, Temperature Coefficient of Reactivity, remains unchanged, but is renumbered to 5.3.3. Temperature Coefficient of Reactivity remains a design feature of the NTR as discussed in criticality safety analysis as it establishes the primary coolant system is not necessary for cooling of the fuel.

Page 20 of 24 GE Nuclear Test Reactor Roadmap of Changes to POL R-33 and Technical Specifications

21. Section 5.4, Fissionable Material Storage Specification 5.4.1, Fuel Storage, has been deleted as the proposed POL R-33 amendment will effectively limit fissionable material at the NTR to the material (fuel) stored in the reactor (See LC 2.8(2) and LC 2.B(2)b above.).

The Basis for section 5.4 has been deleted as the proposed design features are adequately self explanatory.

22. Section 6.1, Organization Figure 6.1.1, Structure, has been revised to reflect new titles and reporting lines. Staffing changes, responsibilities, and associated TSs are discussed in detail in the POL Safety Analysis.

Specification 6.1.2, Responsibilities, has been revised to adjust titles and replace 10 CFR 55 Licensed Operators with Certified Fuel Handlers in preparation for defueling.

6.1.2(2) clarifies that the Level 2 Reactor Administrator has the responsibility of ensuring the NTR is maintained according to the facility license and applicable regulations. While this is implicit in the POL Safety Analysis through the described responsibilities of the Level 2 Reactor, TS 6.1.2(2) clearly states these responsibilities.

6.1.2(3) has been added to address the Radiation Safety Function included in Figure 6-1. The Radiation Safety Function is headed by the Radiation Safety Officer (RSO). This position is required by the site's broad scope SNM-960 and California 0017-01 byproduct material licenses and is a site-wide position that is described in GEH License Amendment Request - Permanent Cessation of the GE Nuclear Test Reactor (NTR) and Possession Only Authorization (ML23279A110), submitted 10/6/2023.

6.1.2(4) and (5) have been renumbered to account for the addition of 6.1.2(3) and edited to clearly state full titles of the Level 3 and Level 4 positions.

6.1.2(6) has been renumbered to account for the addition of 6.1.2(3).

Specification 6.1.3, Staffing, has been edited to delete minimal staffing requirements except that a Certified Fuel Handler Supervisor shall be present in the reactor cell during fuel handling operations.

Specification 6.1.4, Selection and Training of Personnel, has been revised to delete the ANSI/ANS 15.4, 2016, staffing requirements and establish qualifications for the proposed Level 2 Reactor Administrator, Level 3 Certified Fuel Handler Supervisors, and Level 4 Certified Fuel Handlers. This specification also establishes the need for an NRC-approved training program for Certified Fuel Handlers.

6.1.4(1) has been restated to clarify that the Reactor Administrator shall meet the stated minimal standards for the position.

6.1.4(2) has been restated to clarify that the Certified Fuel Handler Supervisors and Certified Fuel Handlers shall be trained according to the NRC-approved training program.

Page 21 of 24 GE Nuclear Test Reactor Roadmap of Changes to POL R-33 and Technical Specifications

23. Section 6.2, Review and Audit Specification 6.2.1, Composition and Qualifications, removes the Regulatory Compliance function and establishes the chartered site Oversight Committee as the sole implementing body for review and audit functions at the NTR. References to the Vallecitos Technological Safety Council (VTSC) have been deleted and replaced with the more generic "Oversight Committee."

Specification 6.2.2, Charter and Rules, has been updated to replace VTSC with Oversight Committee. The reference to NTR operations staff in 6.2.2(3) has been deleted.

Specification 6.2.3(1) has been revised to remove the reference to EXPERIMENTS as experiments are no longer authorized at the NTR.

Specification 6.2.3(2) has been deleted, as experiments are no longer authorized at the NTR.

Specification 6.2.3(3) has been deleted as changes to the Fire Protection program no longer adversely affect the ability to achieve and maintain safe shutdown and no longer fall within applicability of 10 CFR 50.59.

Specification 6.2.3(4), (5), and (6) have been renumbered to 6.2.3(2), (3), and (4) due to deletions.

Specification 6.2.3(7) has been deleted as an internal review of reports made to the NRC in committee does not provide value under the proposed reporting requirements in proposed TS 6.7.2.

Specification 6.2.3(8) has been deleted as the reactor is no longer operational and the inferred reporting criteria (that could affect nuclear safety) is no longer applicable.

Specification 6.2.4, Audit Function, has been edited to remove references to reactor operations and to update titles. Audits are performed under the direction of the Oversight Committee and reports are submitted to the Oversight Committee within 3 months of completion.

6.2.4(2) is revised to replace auditing of the license operator requalification program with an audit of the Certified Fuel Handler Training Program.

6.2.4(3) is revised to remove the requirement that audits be performed of condition reports initiated relative to the operation of the NTR. Audits will continue to be performed of condition reports initiated relative to the NTR.

6.2.4(4) is revised to restrict audits to NTR emergency response procedures, rather than to the site Radiation Emergency Plan.

24. Section 6.3, Radiation Safety This section has been edited to adjust organization titles and responsibilities according to changes made in section 6.1. A reference to reactor operations has been changed to "production organization" to clarify that the reactor is not being operated. Also, the Level 3 Supervisor is currently an optional position. Therefore, "when assigned" has been changed to "in his absence" to indicate that a Level 3 Certified Fuel Handler Supervisor will be assigned for fuel movement.

Page 22 of 24 GE Nuclear Test Reactor Roadmap of Changes to POL R-33 and Technical Specifications

25. Section 6.4, Procedures This section has been edited to reflect changes in organization titles and responsibilities according to changes made in section 6.1. Verbiage has been edited to incorporate section 6.4.2, Level 2 Approval, into the introductory paragraph.

Specification 6.4.1(1) has been deleted as procedures will no longer be necessary for startup, operation, and shutdown of the reactor.

Specification 6.4.1(2) has been edited to remove reference to refueling and fuel transfer operations and this specification has been renumbered to 6.4.1(1). Verbiage has been added to specify that fuel handling is not authorized under the POL prior to defueling and to allow for fuel handling procedures to be prepared in advance but maintained inactive in preparation for defueling.

Specification 6.4.1(3) has been edited to clarify that the fuel remaining in the reactor is "fuel in storage" and is not intended as fuel for reactor operations. Specification 6.4.1(3), 6.4.1(4) and 6.4.1(5) have been renumbered to 6.4.1(2), 6.4.1(3), and 6.4.1(4) to adjust for the deletion of 6.4.1(1).

Specification 6.4.1(6) has been deleted as operation and experiments are not authorized under the POL and to reflect that maintenance cannot affect core reactivity under the POLSC.

Specifications 6.4.1(7) and 6.4.1(8) have been renumbered to 6.4.1(5) and 6.4.1(6) to adjust for the deletion of 6.4.1(1) and 6.4.1(6).

Specification 6.4.2, Level 2 Approval, has been deleted as these responsibilities have been integrated into the section 6.4 introductory paragraph.

Specification 6.4.3, Administrative Changes to Procedures, has been edited to update organizational titles according to changes made in section 6.1 and to specify that the Level 3 Certified Fuel Handler Supervisor, once assigned, has limited authority to make administrative changes to fuel handling procedures. This specification is renumbered to 6.4.2 to adjust for the deletion of 6.4.2.

Specification 6.4.4, Temporary Deviations, has been updated to allow the Level 3 Certified Fuel Handler Supervisor limited authority to deviate from fuel handl ing procedures in order to deal with special or unusual circumstances. This specification is renumbered to 6.4.3 to adjust for the deletion of 6.4.2.

26. Section 6.5, Experiments Review and Approval This section has been revised to state that experiments are no longer performed at the NTR. The balance of section 6.5 has been deleted.

27. Section 6.6, Required Actions The content of this section has been deleted and/ or relocated to section 6. 7, Reports. An introductory statement has been added to clarify that safety limit violations are not applicable under the POLSC as reactor operations are not authorized. Specification 6.6.1 deals with violations of the safety limit, which is not possible under the POLSC. Specification 6.6.2(1) has been deleted as it deals with shutting down the reactor in response to occurrences itemized in specifications 6.7.2(1)b and 6.7.2(1)c. As explained below, specification 6.7.2(1)c has been deleted, leaving only 6.7.2(1)b, reporting a release of Page 23 of 24 GE Nuclear Test Reactor Roadmap of Changes to POL R-33 and Technical Specifications

radioactivity from the site above allowed limits to the NRC and area manager, as the only remnant in section 6.6.2. Notably, reporting to the NRC is already addressed in 6.7.2(1)b. Therefore, immediate reporting to the Level 2 Reactor Administrator has been added to 6.7.2(1)b as per below. As a result, the narrative included in proposed section 6.6. has been edited to state that all content of section 6.6 is either not applicable or has been relocated to section 6.7.2.

28. Section 6. 7, Reports The title for specification 6.7.1, Operating Reports, has been changed to Routine Report and the introductory paragraph for specification 6.7.1 has been rephrased to clarify that the annual report is now required in intervals not to exceed 24 months according to 10 CFR 50.59(d)(2), is no longer a reactor operating report, and that the report is in accordance with the requirements of 10 CFR 50.59.

Specification 6.7.1(1) has been deleted as reactor critical hours and energy produce are no longer applicable under the POL.

Specification 6.7.1(2) has been deleted as unscheduled shutdowns are no longer applicable under the POL.

Specification 6.7.1(3) has been renumbered to 6.7.1(1) to adjust for the deletion of 6.7.1(1) and 6.7.1(2).

Specification 6.7.1(4) has been renumbered to 6.7.1(2) to adjust for the deletion of 6.7.1(1) and 6.7.1(2) and has been edited to reflect the detailed reporting requirements of 10 CFR 50.59(d)(2).

Specification 6.7.1(5) has been deleted as effluents are no longer produced by the NTR. This is due to several reasons addressed in the POL Safety Analysis document, including: reduced operation of the ventilation system, discontinuance of the formation of fission products generated by the operating reactor, and limiting of airborne contamination-generating activities performed under the POL.

Specification 6.7.1(6) and 6.7.1(7) have been renumbered 6.7.1(3) and 6.7.1(4) to adjust for deletion of specifications 6.7.1(1), 6.7.1(2), and 6.7.1(5).

Specification 6.7.2(1)a has been deleted as a violation of safety limit is not applicable under the POLSC.

Specification 6.7.2(1)b has been renumbered to 6.7.2(1)a to adjust for the deletion of 6.7.2(1)a. The requirement that the Level 2 Reactor Administrator shall be immediately notified has been added to allow for deletion of specification 6.6.2.

Proposed specification 6.7.2(1)b has been added to require reporting of abnormal and significant degradation in reactor fuel, cladding, or coolant boundary, which could result in exceeding prescribed radiation limits for personnel or the environment.

Specification 6.7.2(1)c has been deleted as it deals with the reporting of operational violations that are no longer applicable under the POL.

Specification 6.7.2(2)b has been edited to reflect that bounding analysis for the NTR POLSC contained in Chapter 13 of the SAR has been superseded by the criticality analysis in document CSA 008N0128.

Page 24 of 24 GE Nuclear Test Reactor Roadmap of Changes to POL R-33 and Technical Specifications

Proposed specification 6.7.2(3) has been added to provide notification of specific unsafe conditions. This notification is discussed in the POL Safety Analysis.

29. Section 6.8, Records Sections 6.8.1, 6.8.2, and 6.8.3 have been reformatted to "all caps" to set them apart as section titles.

Specification 6.8.1(2) has been edited to replace the word "operations" with "activities" to provide clarity that the reactor is not operational under the POL.

Specification 6.8.1(6) has been edited to remove "all caps" from the word "experiments" as the term is no longer defined in section 1.1. This is in keeping with the format established in revision 6 of the TSs -

that defined terms used in the body of the TSs are capitalized and italicized.

Attachment 3 NTR Certified Fuel Handler Training and Requalification Program (CFHTRP)

Dated January 2024 Page 1 of 3 January 2024

CERTIFIED FUEL HANDLER TRAINING PROGRAM FOR THE NUCLEAR TEST REACTOR Pa ge 1 of3

1. OBJECTIVE

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.

2. EXPERIENCE/QUALIFICATIONS

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 ).

3. ROLES/RESPONSIBILITIES 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.

Pag e 2 of 3 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.

4. TRAINING PROGRAM

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.

Page 3 of3 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 basis 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.

5. ACTIVE/INACTIVE STATUS

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 and 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.

6. CFH HEAL TH QUESTIONNAIRE AND REVIEW

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.

7. RECORDS

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.

v t e Letter Sequence Response to RAI
EPID:L-2023-LLA-0139, Enclosure 4: Nedo 32740, Revision 5, General Electric Nuclear Test Reactor Safety Report (Public) (Approved, Closed)
Initiation Request, Request Acceptance... Administration Withholding Request Acceptance Questions 1 November 2023 Answers 15 December 2023 26 January 2024 Results Approval Other: ML23027A211, ML24031A123
MONTHYEAR2023-12-15 [Table View]

ML24026A196

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