ML23349A047

GEH Response to NRC Request for Additional Information

ML23349A047
Person / Time
Site:	Vallecitos Nuclear Center
Issue date:	12/15/2023
From:	Murray S GE Hitachi Nuclear Energy
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
M230143
Download: ML23349A047 (1)

Text

GE Hitachi Nuclear Energy

• HITACHI Scott P. Murray

Manager, Facility Licensing

3901 Castle Hayne Road P.O. Box 780 December 15, 2023 Wilmington, NC 28402 USA M230143 T (910) 819-5950 scot!. 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 (M L23298A 154)

9) Letter, G. Wertz (NRC) to C. Martinez (GEH), Request for Additional Information, 11/1/23 (ML23305A051)

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 GEH 's Vallecitos Nuclear Center (VNC) NRC license R-33 in Sunol, CA.

Attached to this letter are several of the documents requested by the NRC (References 8 and 9).

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

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

s* ly, i.1111 t.;.:z5G I I I

• urray, Manager~

Facility Licensing

• M230143 U.S. NRC December 15, 2023 Page 2 of 2

Redacted (public) version of Enclosure 3 - Criticality Safety Analysis (008N0128), Revision 0, dated September 2023 : GEH Responses to Request for Additional Information (RAI) 1 and 2 in the NRC Request dated November 1, 2023

Cc: G. Wertz, USNRC/NRR/DANU/UNPL SPM 23-040 Attachment 1 Enclosure 3 GEH Possession Only License Amendment Safety Analysis of the NTR (Part 2)

Contains Proprietary Informat ion Gonfidentia{ Pfoprietar1,'-nfofmation V!/ithhold from DisO:osure Pufsuant to 10 GFR 2.390

- HITACHI GE Hitachi Nuclear Energy

008N0128 Revision O September 2023

GEH PFe13FietaFy IA:feFffiatieA ~JeA Puslie

GEH Possession Only License Amendment Safety Analysis of the NTR

Copyright 2023 GE-Hitachi Nuclear Energy Americas LLC, All Rights Reserved Gonfidentia{ Pfoprietar1,'-nfofmation V!/ithhold from DisO:osure Pufsuant to 10 GFR 2.390 008N0128 REVISION 0

PROPRIETARY INFORMATION NOTICE

HliS elocuA9eAt COAtaiAS pFOpFietary iAfoFA9atiOA of GE Hitachi ~JucleaF EAeFgy /\\A9eFicas LLG (GEH) aAd is fUFAisheel iA COAfieleAce solely foF the puFpose(s) stateel below iA the AOtice FCgaFeliAg the COAteAtS of this Feport. No otheF use, diFCet OF iAeliFeet, of the docuA9eAt OF the iAfOFA9atiOA it COAtaiAS is authoFiz:ed.

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IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT

Please Read Carefully

The design, engineering, and other information contained in this document are furnished for the purpose of supporting the Nuclear Regulatory Commission (NRC) review and approval of the Possession Only License for the GEH Nuclear Test Reactor (NTR). The use of this information by anyone other than the NRC, or for any purpose other than that for which it is furnished by GEH is not author ized; and with respect to any unauthorized use, GEH makes no representation or warranty, express or implied, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document, or that its use may not infringe privately owned rights.

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REVISION

SUMMARY

Rev# Section Revision Summary Modified 0 - Initia l Issue

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TABLE OF CONTENTS

1.0 INTRODUCTION

.............................................................................................................. 6 2.0 METHODOLOGY............................................................................................................. 7 2.1 NTR Description...................................................................................................... 7 2.2 POL Shutdown Configuration.................................................................................. 7 2.3 HAZOP Analysis and AOOs.................................................................................... 8 2.4 NTR Core Computational Model.............................................................................. 8 3.0 INPUTS.......................................................................................................................... 10 3.1 Evaluation Basis..................................................................................................... 1 O 3.2 Materials................................................................................................................. 10 3.3 Source Geometry................................................................................................... 11 3.4 Criticality Safety Assessment.................................................................................. 13 3.4.1 Movable Neutron Poison Reactivity Worth.................................................. 13 3.4.2 Moderator Reactivity Worth......................................................................... 13 3.4.3 Testing Facility and Fuel Chute Graphite Reactivity Worth.......................... 14 4.0 RESULTS....................................................................................................................... 15 4.1 Core Computational Model Scenario...................................................................... 15 4.2 Critical Eigenvalue Comparison.............................................................................. 16

5.0 CONCLUSION

................................................................................................................ 18 5.1 Criticality Safety Assessment Summary................................................................. 18 5.2 Shutdown Configuration Restraining Efforts and Surveillance................................ 18 5.2.1 Restraining Efforts of Neutron Poisons........................................................ 18 5.2.2 Continuing Surveillance Actions.................................................................. 18 6.0 ACRONYMS AND DEFINITIONS................................................................................... 20 6.1 Acronyms............................................................................................................... 20 6.2 Definitions............................................................................................................... 20

7.0 REFERENCES

............................................................................................................... 21

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LIST OF TABLES Table 3-1: NTR Material Descriptions...................................................................................... 11 Table 3-2: MCNP Fuel Assembly Model Dimensions............................................................... 12 Table 4-1: MCNP Case Descriptions........................................................................................ 15 Table 4-2: Reactivity Worth Results......................................................................................... 16 Table 4-3: Reactivity Subcriticality Results............................................................................... 17

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1.0 INTRODUCTION

This evaluation details the work completed to support the transition of the Nuclear Test Reactor (NTR) R-33 license [1] to a Possession Only License (POL). This evaluation describes and supports the subcritical configuration of the NTR pending decommissioning. The fuel will remain inside the reactor unt il further notice. The NTR Safety Analysis Report (SAR) [2] and Technical Specifications (TS) [3] will be updated to support the baseline subcritical configuration for the POL and detail the necessary surveillance and maintenance procedures.

The previous MCNP model in the latest SAR revision, which was reviewed by the NRC, is used for this analysis. The baseline subcritical configuration is perturbed based on credible accident scenarios to perform a Criticality Safety Analysis (CSA) of the POL configuration of the NTR.

NUREG - 1537 Part 1 Section 17.2 details the guidance for a POL and serves as the framework for this CSA [4]. The safety analysis requ ired must show that the fac ility can be possessed in a way that protects the health and safety of workers, the pub lic, and the environment. The spectrum of credible accident scenarios for a POL safety analysis is limited to those for a reacto r in a shutdown configuration and for the residual rad ioactive material not described in the SAR for the operating fac ility [1 ].

This criticality safety analysis (CSA 008N0 128) provides support fo r, but does not constitute the entirety of, the POL safety analysis required by NUREG-1537.

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2.0 METHODOLOGY

2.1 NTR Description

The NTR is located at the Vallecitos Nuclear Center (VNC) Site and was designed to perform neutron radiography, small sample irradiation, sensitive reactivity measurements, training, calibrations, and other testing utilizing a neutron flux [2,3). The NTR is a heterogeneous, high enriched-uranium, graphite moderated and reflected, light-water cooled, thermal reactor. It does not operate at power levels greater than 100 kW. The fuel consists of high enriched uranium aluminum alloy disks, clad with aluminum. The NTR core consists of an aluminum can filled with water in a graphite pack. The core is cooled either by natural or forced circulation of deionized light-water circulated in a primary system constructed primarily of aluminum. The reactor operates at a very low temperature and low heat flux. Reactivity is controlled by up to six manually positioned cadmium sheets, four boron-carbide-filled safety rods, and three electric-motor-driven boron-carbide-filled control rods.

2.2 POL Shutdown Configuration

For the POL configuration, the NTR fuel will remain in the reactor within the aluminum can described in Section 2.1. The baseline configuration is then perturbed according to credible accident scenarios from the SAR or scenarios determined by an additional hazard screening. This collection of possible scenarios that could affect the baseline configuration are analyzed to demonstrate that the system remains subcritical.

The baseline subcritical configuration consists of the safety rods and control rods fully inserted into the NTR core. This is in contrast to the operational subcritical rod position, which requires that only the control rods be inserted. Once inserted, the safety and control rods will be secured rendering the drives incapable of withdrawing the rods. The drive chain for the coarse control rods, CR-1 and CR-2, will be removed, and they will be electrically isolated. The fine control rod will be electrically isolated. The drive belts will be removed from all four safety rods, and they will be electrically isolated.

The Manual Poison Sheets (MPS) contain cadmium poison between two aluminum plates. Three of the six MPS positions are equipped with latches that provide positive restraint with spring loaded latch handles on the sheets [2]. Only three sheets are used in the baseline configuration with the following widths in the three available locations:

• Position 1, 2.5" - This is equivalent to 9/10 MPS sheet.

• Position 2, 1.375" - This is equivalent to1/2 MPS sheet.

• Position 5, 2.75 " - This is equivalent to a full MPS sheet.

The MPS will be restrained in the baseline configuration using locks that will be placed on the position covers. The keys will be removed from the facility.

These configuration updates will be documented in Section 7.5 of the SAR as the new system description.

The core tank will be full of water. The primary coolant system, discussed in Section 5.2 of the SAR, will continue to flow so that water chemistry can be maintained through the demineralizer cartridges.

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As part of this CSA, credible accidents are evaluated from the baseline configuration following the completion of a Hazards and Operability Analysis (HAZOP), detailed in Section 2.3. The changes to the control rods, safety rods, MPS, moderator, and geometry are discussed in Section 3.4.1.

2.3 HAZOP Analysis and AOOs

A HAZOP is performed to determine the credible accident scenarios for the NTR in a shutdown configuration. The purpose of the HAZOP is to show that no single failure will lead to inadvertent criticality based on the events evaluated. Possible hazards are evaluated based on natural phenomena and failure of administrative controls after identifying and gathering existing procedures, facility and process descriptions, and existing safety documentation.

Following the HAZOP facility walkdown, the following systems are identified to be at risk of having a single mode of failure: movable poisons (MPS, safety rods, control rods), primary coolant loop, and test facility geometry changes. Additionally, the SAR lists the Anticipated Ope rational Occurrences (AOOs) as loss of normal electrical power, loss of facility air supply, loss of secondary coolant, inadvertent core inlet temperatu re changes, and fuel handling errors [2]. The systems identified in the HAZOP a re discussed further in Section 3.4, and the AOOs stated in the SAR are also selected for further qualitative or quantitative investigation.

The NTR facility does not require an emergency power supply. When the reactor is shut down and the safety and control rods are inserted, subcriticality is maintained independent of normal electrical power. Section 1.8.4 of the SAR discusses a 2003 modification that allows the control rods to be bottomed using alternate (non-normal) power. This modification will be disabled in the baseline shutdown configuration. Loss of facility air supply does not affect systems relevant to criticality, so it is not considered in the evaluation. Loss of secondary coolant would result in a tempe rature increase for the primary coolant. The primary reactor coolant has a negative temperature coefficient at 124 °F (See Section 3.1 Item 3), mean ing the reactivity will turn over and decrease when it reaches that temperature. Therefore, no reliance on cooling is needed to ma intain the shutdown condition. This is validated in the evaluation alongside the inadvertent core inlet tempe rature change and discussed further in Section 3.4.2.2.

Per the SAR, the only credible external event fo r the NTR is a seismic event [2]. Various scenarios are modeled in this CSA with different combinations of rods inserted, including a scenario where all rods are removed. A seismic event is also considered in the movement of the MPS. The sheets are mechanically restrained within their slots so that they will not move relative to the core during a seismic event. Despite these restraints, conservative configurations w ith only one or two MPS inserted into the NTR are modeled to demonstrate subcriticality.

Fuel handl ing hazards are not considered. The NRC license for Special Nuclear Material at VNC (SNM - 960) does not author ize the NTR fuel to be stored at either the hillside storage facility or the onsite hot cell fac ility [5,6]. The baseline shutdown configu ration assumes that the NTR fuel will not be removed from the reactor until it is be ing prepared for offsite shipment for bur ial.

2.4 NTR Core Computational Model

This analysis is completed w ith MCNP-06P, which is the GEH Level 2 controlled version of the Los Alamos National Laboratory (LANL) code MCNP6 Version 2 (MCNP). The ENDF/8-VII cross section library is used fo r the full-core model criticality and neutronics calculat ions.

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MCNP6 uses pointwise (i.e., continuous) cross section data, and all reactions in a given cross section evaluation are considered. In this evaluation, thermal neutron scattering with hydrogen and with graphite are described using an S(a,13) thermal scattering kernel for light water and graphite, respectively.

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3.0 INPUTS

An existing verified NTR axe computational model approved by the NRC for the purposes of relicensing is used as the foundation for the model input files [1]. The full core MCNP model is generated using available information on the 16 fuel assemblies and surrounding systems.

Additional model details are described in Sections 3 2 and 3.3.

3.1 Evaluation Basis

1 The current NTR exposure statepoint used is the value from 9/30/2019. Conservat ively, no credit is taken for additional fuel burnup since 2019 (Le, the fuel is modeled as more reactive than current fue l exposure conditions).

2. The verified and NRC approved NTR Core Computational Model is used as the basis for this evaluation [2].

3. The water temperature and void reactivity feedback is based on the primary coolant temperature coefficient of reactivity relationship in SAR Equation 4 - 1 [2]. The overall temperature coefficient of the fuel annu lus is positive up to and at 124 °F. Then, as water temperature is increased above 124 °F (the turnover point), the p rimary coolant temperature coefficient of reactivity becomes negative.

4 _ The system average r,3ett value is [ l [2}.

5 _ The average critical eigenvalue for the NTR from the model benchmarking is [

which was used to support the relicensing effort [1, 2].

6_ Only the water temperature is assumed to increase in the event of a secondary coolant failure. The temperature effects of the graph ite are not considered [2J.

3.2 Materials

The NTR MCNP model cell description, material description, and density are included in Table 3-1.

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Table 3-1: NTR Material Descriptions

Cell Description Material Description Material Density [g/cm3]

Fuel Uranium Alloy 3.032 Rod and Cladding 1100 series Aluminum 2.70 Coolant Light Water 1.0 Cannister Shell, Fuel Loading 1100 series Aluminum 2.70 Chute, and Flow Distributors Poison Sheet Slots and 6061-T6 Aluminum 2.70 Material MPS Cadmium MPS Absorber - Pure Cadmium 8.65 Safety Rods and Control Rods T300 Stainless Steel 7.89 Cladding Safety Rods and Control Rods Boron Carbide 1.512 Absorber

Fue l Loading Chute Plug Moderator - AGOT Nuclear 1.65 Graphite

Source Log - Plastic Disks PVT Plastic 1.032 Source Log - Lead Disks Lead 11.35 Source Log - Assembly, Outer Tube, End Caps, Aluminum 6061-T6 Aluminum 2.70 Disks Source Log - Graphite AGOT Nuclear Graphite 1.65 Cylinder Main Graphite Pack and AGOT Nuclear Graphite 1.65 Thermal Column Horizontal Facility and AGOT Nuclear Graphite 1.65 Graphite Sleeve Within Cable Held Retractable Irradiation System (CHRIS) 1100 series Aluminum 2.70 Facility, Tube Walls

3.3 Source Geometry

The fuel disks in the MCNP model are stacked so that the fuel assembly active length is in the x-direction. Each assembly consists of 40 fuel disks stacked along an aluminum rod and there are 16 fuel assemblies arranged circumferent ially inside the core canister to make up the NTR core. Core modeling details are listed and defined in Table 3-2.

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Table 3-2: MCNP Fuel Assembly Model Dimensions

NTR Component/Fuel Model Model Element Description (cm) Model Parameter (Direction) Dimension Dimension (in)

Inner Diameter (Y, Z-directions; 1.31064 0.516 parallel to X-ax is)

Inner Edge Ring (Inside the Outer Diameter (parallel to X-axis) 1.47828 0.582 Inner Diameter of the Fuel to X- 0.08382 Disk) Radial Thickness (parallel axis) 0.033

Height (X-direction) 0.36068 0.142 Inner Diameter (Y, Z-directions; 1.47828 0.582 parallel to X-ax is)

Uranium-Aluminum Alloy Fuel Outer Diameter (parallel to X -axis) 6.8072 2.68 Disk Fuel Disk Height (X-direction) 0.22352 0.088

Aluminum Disk Height (X- 0.06858 0.027 direction)

Inner Diameter (parallel to X-axis) 6.8072 2.68 Outer Edge Ring Outer Diameter (parallel to X-axis) 7.04339 2.773 (Surrounding the Outer Radial Thickness (parallel to X- 0. 1181 0.0465 Diameter of the Fuel Disk) axis)

Height (X-direction) 0.36068 0.142 Distance 1 Face-to-Face Spac ing between Uranium-Aluminum Alloy 0.60452 0.238 Fuel Disks (X-direction)

Distance 2 Face-to-Face Spacing between Uranium-Aluminum Alloy 0.68326 0.269 Fuel Disks (X-direction)

Aluminum Spacer He ight, Active Fuel Length included in Distances 1 and 2 0.4572 0.180 Face-to-Face Spac ing between Fuel Disks (X-direction)

Aluminum Washer Height, included in Distance 2 Face-to- 0.07874 0.031 Face Spacing between Fuel Disks (X-direction)

Active Fuel Length ( X-direction) 39.49954 15. 551

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3.4 Criticality Safety Assessment

3.4.1 Movable Neutron Poison Reactivity Worth

There are three types of movable neutron poisons in the NTR to control core reactivity [2}. These are the MPS, safety rods, and control rods. These elements will be secured in place in the baseline configuration, as discussed in Section 2.2.

3.4.1.1 Manual Poison Sheets

There are three MPS slots that are available for the sheets located within the graphite reflector

[2}. These are located in positions 1, 2, and 5. The MPS are only able to be moved by entering the reactor cell and removing a shield plug in the shield face. The three sheets listed in Section 2.2 will be installed for the POL shutdown configuration.

To evaluate reactivity worths, cases are run with varying MPS slots occupied. Additionally, one case is run with no MPS sheets, but all safety and control rods inserted. Based on previous benchmarking, the limiting MPS slot is in position 5. For the cases with one or two MPS sheets in place, the locations are determined based on MPS width available, removing the largest and second largest sheets to demonstrate the most conservative criticality evaluations.

3.4.1.2 Safety and Control Rods

There are four safety rods, two coarse control rods, and one fine control rod. The poison section of each safety rod is 20 inches long and has a 1/2-inch diameter core of solid boron carbide within a stainless steel tube. The poison section of the coarse control rods is 16 inches long with a solid 1/2-inch diameter boron carbide core contained in a stainless steel tube. The poison section of the fine control rod is 18 inches long and has a solid 0.365-inch diameter core of boron carbide contained in a stainless steel tube [2].

Two independent cases are evaluated with the highest worth coarse control rod (CR-2) and highest worth safety rod (SR-2) removed. The reactivity worths of the individual rods are listed in SAR Section 4.2.2 [2]. A case with all rods withdrawn is also evaluated. These control rod and safety rod selections demonstrate the subcriticality of a significantly conservative configuration.

3.4.2 Moderator Reactivity Worth

3.4.2.1 Flooding

The flooding cases evaluated are the filling of the available MPS slots with water and filling the horizontal test facility and fuel chute plug with water.

When no MPS sheets are inserted in the NTR, the MPS slots are modelled as a void. The empty space in these slots is moderated with water to observe the effects on criticality. The horizontal facility and fuel chute contain removable graphite slugs. The effect on criticality of removing these slugs and moderating the region is considered.

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3.4.2.2 Primary Coolant Temperature

The typical core inlet temperature is 90 °F and outlet temperature is 124 °F with a flow rate of 20 gpm [2]. To evaluate the reactivity worth of the primary coolant temperature increase, the primary coolant temperature is raised from a conservative starting temperature of 75 °F to 124 °F. This reactivity evaluation did not require an MCNP run. Instead, the following equation from the SAR is used [2]:

6.p(T) = 2.85 X 10- 3 X (T - 124) 2 [¢] (3-1)

Where:

8.p = The change in reactivity, in cents T = Coolant temperature, in °F This equation is evaluated at the primary coolant temperature of 75 °F to determine the reactivity added from increasing the primary coolant temperature from 75 °F to 124 °F.

3.4.3 Testing Facility and Fuel Chute Graphite Reactivity Worth

The vertical facility is a 4-inch by 5-foot-long aluminum can that extends from the top to the bottom of the reflector within the graphite pack [2]. The horizontal facility is a 5-inch diameter hole through the horizontal axis of the reactor. The fuel chute is a rectangular aluminum chute approximately 30 inches long, 20 inches wide, and 3 inches high. These facilities are filled with graphite when not being used for testing. One case is run analyzing the effects of removing the graphite from the vertical facility. A second case is run analyzing the combined effects of removing the horizontal facility and fuel chute plug graphite blocks. These cases encompass the various testing configurations used at the NTR and demonstrate the reactivity effects of removing the graphite blocks.

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4.0 RESUL TS 4.1 Core Computational Model Sce n ario

A summary table of the cases analyzed with MCNP are detailed in Table 4-1.

T able 4-1: MCNP Cas e Descriptions

Case MPS Pos ition Saf ety Ro d Co n trol Ro d Addition al Notes Numbe r Po sitio n Pos ition 9/10 MPS in P1, 1/2 1* MPS in P2, full MPS Fully Inserted Fully Inserted in PS

2 9/10 MPS in P1, 1/2 Fully Inserted Fully Inserted MPS in P2

3 1/2MPSinP2 Fully Inserted Fully Inserted 4 NIA Fully Inserted Fully Inserted 9/10 MPS in P1, 1/2 s MPS in P2, full MPS SR-2 out Fully Inserted in PS 9/10 MPS in P1, 1/2 6 MPS in P2, full MPS Fully Inserted CR-2 out in PS 9/10 MPS in P1, 1/2 Fully 7 MPS in P2, full MPS Fully Withdrawn in PS Withdrawn 8 N/A Fully Inserted Fully Inserted MPS space flooded 9/10 MPS in P1, 1/2 Horizontal facility and fuel 9 MPS in P2, full MPS Fully Inserted Fully Inserted in PS chute flooded 9/10 MPS in P1, 1/2 Horizontal facility and fuel 10 MPS in P2, full MPS Fully Inserted Fully Inserted chute graphite plug in PS removed 9/10 MPS in P1, 1/2 Fully Vertical facility graphite 11 MPS in P2, full MPS Fully Withdrawn in PS Withdrawn plug removed 9/10 MPS in P1, 1/2 Primary coolant 12 MPS in P2, full MPS Fully Inserted Fully Inserted temperature increased from in PS 75 O f to 124 O f

• Th is is the baseline configuration.

• No MCNP case was completed for this case. The k-eff is estimated based on the reactivity change from Eq. 3-1 from the baseline configuration.

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Table 4-2 lists the reactivity worths of the cases listed above based on the discussed baseline configuration changes in Section 3.4. These results include the eigenvalue plus two standard deviations to account for computational bias and bias uncertainty. These values are converted to reactivity in units of dollars based on the average r,3ett value, using the baseline configuration results as the initial value.

Table 4-2: Reactivity Worth Results

Case Number k@n 0 kerr+2o Reactivity ($)

1 0.92707 0.000 13 0.92733 0.00 2 0.93840 0.00011 0.93862 +1.55 3 0.94845 0.00011 0.94867 +2.94 4 0.95540 0.00012 0.95564 +3.89 5 0.93248 0.00012 0.93272 +0.74 6 0.93332 0.00012 0.93356 +0.86 7 0.96852 0.00011 0.96874 +5.70 8 0.95421 0.00011 0.95443 +3.73 9 0.88539 0.00011 0.88561 -5.74 10 0.91363 0.00012 0.91387 -1.85 11 0.96179 0.00012 0.96203 +4.77 12+ 0.92757 +0.07

• No MCNP case was completed for !his case. The k-eff is estimated based on the reactivi1y change from Eq. 3-1 from the baseline configuration.

4.2 Critical Eigenvalue Comparison

The NTR core model average critical eigenvalue is ] based on previous benchmarking, which was used to support the relicensing effort [1]. The cases evaluated in this analysis demonstrate a level of subcriticality based on the difference from this critical eigenvalue. The results of this comparison are shown in Table 4-3.

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Table 4-3: Reactivity Subcriticality Results

Case Number ken a kett+2o Margin to Critical ($)

1 0.92707 0.00013 0.92733 +9.82 2 0.93840 0.00011 0.93862 +8.27 3 0.94845 0.00011 0.94867 +6.88 4 0.95540 0.00012 0.95564 +5.92 5 0.93248 0.00012 0.93272 +9.08 6 0.93332 0.00012 0.93356 +8.96 7 0.96852 0.00011 0.96874 +4.12 8 0.95421 0.00011 0.95443 +6.09 9 0.86539 0.00011 0.88561 +15.56 10 0.91363 0.00012 0.91387 +11.67 11 0.96179 0.00012 0.96203 +5.05 12* 0.92757 +9.79

• No MCNP case was com pleted for this case. The k-eff is estimated based o n the reactivity change from Eq. 3-1 from the base line configuration.

Case 7 has the lowest marg in to the critica l eigenva lue. Cons idering that this de lta is greater than the NTR ope rating license shutdown ma rgin of +$2.0 [2), there is adequate marg in to maintain subcritica lity in the wo rst-case configu ration w ith a ll safety and control rods ejected.

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5.0 CONCLUSION

5.1 Criticality Safety Assessment Summary

Scenarios perturbing the baseline shutdown configuration described in Section 2.2 are evaluated to support a POL for the NTR decommissioning efforts. The results in Section 4.0 show sufficient subcritical margin in each of the credible accident scenarios. Having the NTR remain shut down in the baseline configuration meets the safety standards set forth.

5.2 Shutdown Configuration Restraining Efforts and Surveillance

Securing the NTR fuel in place will require continuing maintenance and surveillance to support a POL. To prevent inadvertent reactivity changes, the movable poisons will be restrained. The following sections outline the changes and maintenance needed to support the preventative and mitigative controls for the systems identified in the HAZOP. These requirements will be captured in the SAR and technical specifications.

5.2.1 Restraining Efforts of Neutron Poisons

To maintain the baseline shutdown configuration discussed Section 2.2, there are several restraining efforts that need to be implemented to prevent inadvertent scenarios that could lead to an increase in reactivity.

1. MPS Slots - Locks will be placed on the MPS slot covers and the keys will be removed from the facility.

2. Control rods - The drive chains will be removed from CR-1 and CR-2, and they will be electrically isolated. The fine control rod will be also electrically isolated.

3. Safety Rods - The drive belt will be removed from all four safety rods, and they will be electrically isolated.

5.2.2 Continuing Surveillance Actions

There are several systems that will require continuing surveillance while the NTR is in the baseline shutdown configuration.

The stack gas effluent gaseous and particulate activity systems are described in Section 11. 1.4 of the SAR. These detect gaseous and particulates from the stack and are located in the NTR setup room closet. There is a Hoffman blower that provides the motive force for air sampling. It is located in the Auxiliary Machinery Space (AMS) of Bldg. 105. There is a readout chart recorder with alarm function in the control room.

The water chemistry system demineralizer cartridges, as described in Section 5.4 of the SAR, are located in the reactor cell and are replaced annually. Additionally, there are demineralizer cartridges located in the AMS that are replaced as needed. This system requires the primary coolant pump to be online because the differential pressure drives the flow through the cleanup system. There is a mechanical filter attached to this component and is replaced annually. The conductivity monitor is located in the control room and is calibrated biennially.

The sump pump, described in Chapter 5 of the SAR, is located in the northwest corner of the reactor cell. It is flooded and checked annually for operation.

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Level in the primary system is monitored as described in Section 5.5 of the SAR. The low tank level alarm described therein will remain in service. There is a primary coolant valve, numbered V-140 (see SAR Figure 5-1 ), that is maintained in the instance that water needs to be added or removed from the primary coolant loop. This valve must remain operable.

Functional area radiation monitors, or equivalent, will need to be maintained to monitor the reactor cell.

The minimum shutdown margin is evaluated biennially per procedure by showing the worths of rods by moving them. These systems will be assessed prior to the restraining of the systems for the baseline shutdown configuration. The conditions of the shutdown configuration are such that the systems will not lead to any degradation or corrosion. The control rods, safety rods, and MPS will not be moved, therefore no boron leakage is expected consequential to wear-related deterioration of reactor internals. The previously discussed reactivity effects and the shutdown margin will be maintained by the geometric configuration of the NTR and surveillance of equipment status, not operation.

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6.0 ACRONYMS AND DEFINITIONS

6.1 Acronyms

Acronym Explanation

AMS Auxiliary Machinery Space

AOOs Anticipated Operational Occurrences

CHRIS Cable Held Retractable Irradiation System

CSA Criticality Safety Assessment

HAZOP Hazards and Operability Analysis

LANL Los Alamos National Laboratory

MPS Manual Poison Sheets

NRC Nuclear Regulatory Commission

NTR Nuclear Test Reactor

POL Possession Only License

SAR Safety Analysis Report

SNM Special Nuclear Material

TS Technical Specifications

VNC Vallecitos Nuclear Center

6.2 Definitions

Term Definition Reactivity ($) Reactivity in units of$ is the normalized value of reactivity, p, to the delayed neutron fraction, f3err.

The formula for reactivity in $ is dollars = p/f3ett Delayed neutron The fraction of delayed neutrons in the core at creation of high energies.

fraction (f3err)

Critical eigenvalue The system specific eigenvalue, k e11, that results in criticality.

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7.0 REFERENCES

1. NRC Memorandum, GE-HITACHI NUCLEAR ENERGY AMERICAS, LLC-ISSUANCE OF RENEWED FACILITY OPERA TING LICENSE NO. R-33 FOR THE NUCLEAR TEST REACTOR (EPID L-2020-RNW-0038), June 29, 2023 (ADAMS Accession Number ML23128A348).

2. NEDO-32740, General Electric Nuclear Test Reactor Safety Analysis Report, GE Hitachi Nuclear Energy, Revision 5, March 2023.

3. NEDO-32765, Technical Specification for the Nuclear Test Reactor Facility License R-33, GE Hitachi Nuclear Energy, Revision 6, March 2023.

4. NUREG-1537 Part 1, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors, U.S. Nuclear Regulatory Commission, February 1996.

5. NRG Confirmatory Order EA-14-144, April 22, 2015 (ADAMS Accession Number ML14269A172).

6. NRC License SNM-960, Amendment 2, April 20, 2022 (ADAMS Accession Number ML22108A301 ).

GEi-i PFOpriotary lnforfflation Non Publio Pago 21 of 21 Attachment 2 GEH Responses to Request for Additional Information (RAI) 1 and 2 in the NRC Request dated November 1, 2023 M230143 U.S. NRC December 15, 2023 Page 1 of 2 RAil:

The regulations in 10 CFR Part5 70 provide requirements for the licensing of possession of SNM To reflect the permanent cessation ofNTR operations, the NRC staff provides the following suggested POL conditions (italicized text below).

2.B Subject to the conditions and requirements incorporated herein, the Commission hereby licenses GE-Hitachi Nuclear Energy Americas, LLC:

(1) Pursuant to Section 104c of the Act and 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities," to possess the reactor at the designated location in Alameda County, California, in accordance with the procedures and limitations described in the application and set forth in this license.

(2) Pursuant to the Act and 10 CFR Part 70, to possess but not separate:

a. up to 4 kilograms of contained uranium-235 of an enrichment of 20 percent or greater in the isotope uranium-235, in the form of in-core reactor fuel; and

b. such special nuclear material as may have been produced by the operation of the reactor.

(3) Pursuant to the Act and 10 CFR Part 30, to possess:

a. up to 0.2-curie radium-beryllium sealed startup source, and up to 0.03 curies of byproduct materials, in the form of sealed sources, for instrument calibration and source checks; and

b. but not to separate such byproduct material as may have been produced by the operation of the reactor.

C. This license shall be deemed to contain and is subject to the conditions specified in 10 CFR Parts 20, 30, 50, 51, 55, 70, and 73 of the Commission's regulations; is subject to all applicable provisions of the Act, and to the rules, regulations, and orders of the Commission now, or hereafter in effect, and to the additional conditions specified below:

(1) Maximum Power Level The licensee is not authorized to operate the reactor at any power level.

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised by Amendment No.

26, are hereby incorporated in their entirety in the license. The licensee shall maintain the facility in accordance with the Technical Specifications.

M230143 U.S. NRC December 15, 2023 Page 2 of 2 Provide confirmation of acceptability and/or suggested edits to these proposed license conditions.

GEH Response to RAJ 1:

GEH agrees with and accepts these license condition recommendations in totality and has incorporated them in their entirety into a track markup draft to be provided to the NRC in a supplement to the license amendment request to the NRC to modify Facility Operating License No. R-33 for the Nuclear Test Reactor (NTR) to a possession-only license, dated October 6, 2023 (Agencywide Documents Access and Management System Accession No. ML23279Al 10)

RAJ2:

The regulations in 10 CFR 50.82(b)(6)(i) state that the NRC will terminate a non-power reactor license if, in part, it determines that the decommissioning of the facility has been performed in accordance with the approved decommissioning plan (DP).

The guidance in NUREG-1537, Part 1, Section 17.2.2 states that possession-only LARs should discuss the activities to be accomplished and their schedule while the POL is in effect.

It is not clear to the NRC staff what types of activities GEH may plan to conduct following the issuance of a POL, but prior to the NRC approval of an NTR DP Additionally, it is not clear how GEH plans to control activities before DP approval to make sure that no activities are conducted ahead of DP approval that should only be conducted in accordance with an approved DP 2.1. Describe the types of activities that are decommissioning-related or that are done to prepare for decommissioning (e.g., facility characterization activities) that GEH may plan to conduct prior to NRC approval of an NTR DP 2.2 Describe how GEH plans to control activities before the NTR DP approval to make sure that only appropriate activities are conducted. Describe NTR staff responsibility for controlling decommissioning-related activities.

GEH Response to RAJ 2:

1. Several decommissioning-related activities are planned; including but not limited to: Surveillance and Maintenance of the NTR facility, facility and environmental characterizations, development of relevant decommissioning and waste management procedures, and remove non-essential items and wastes that are not directly related to decommissioning the reactor assembly.

2. The Level 1 License Holder will be responsible for controlling the NTR activities before and after NRC approval of the Decommissioning Plan.