ML23261A594

GE-Hitachi Nuclear Energy Americas, LLC, Vallecitos Boiling Water Reactor (Vbwr) License Termination Plan

ML23261A594
Person / Time
Site:	Vallecitos Nuclear Center, Vallecitos  File:GEH Hitachi icon.png
Issue date:	08/18/2023
From:	GE-Hitachi Nuclear Energy Americas
To:	Office of Nuclear Material Safety and Safeguards
Shared Package
ML23261A591	List: ML23261A592 ML23261A593 ML23261A594 ML23261A595 ML23261A596
References
M230128
Download: ML23261A594 (1)

Text

Enclosure 2 VBWR License Termination Plan

e HITACHI License Termination Plan for Vallecitos Boiling Water Reactor (VBWR)

GE Vallecitos Nuclear Center, Sunol, CA

TABLE OF CONTENTS SECTION PAGE

1.0 INTRODUCTION

AND PURPOSE ... .... .... .... .... .... .... ... .... .... .... .... .... .... .... .... .... .... .... .... 1-1 1.1 Site Licensing ...................................................................... .... .... .... .... .... .... .... 1-1 1.2 Transfer of VBWR Residual Contamination to EVESR .................................... 1-2 1.3 Site Location and Description ....... .... .... .... .... .... .... .... ... .... .... .... .... .... .... .... .... .... 1-3 1.4 Overview of LTP Sections ................................................................................ 1-4 2.0 CHARACTERIZATION ............. .... .... .... .... .... .... .... .... ... .... .... .... .... .... .... .... .... .... .... ........ 2-1 2.1 Site Characterization ............. .... .... .... .... .... .... .... .... .... ... .... .... .... .... .... .... .... ........ 2-1 2.2 Historical Site Assessment ............................................................................. 2-12 2.3 Characterization ................................................................................... .... ...... 2-15 2.4 Estimate of VBWR Radiological Inventory for Transfer to EVESR ................. 2-21 3.0 REMAINING D&D ACTIVITIES ................................................................................... 3-1 3.1 Introduction .................................................................. .... .... .... .... .... .... .... ........ 3-1 3.2 Completed Decommissioning Activities & Tasks .............................................. 3-2 3.3 Future Decommissioning Activities .................................................................. 3-3 3.4 Estimate of Quantity of Radioactive Material to be Shipped for Disposal ......... 3-4 4.0 SITE REMEDIATION PLAN ........................................................................................ 4-1 4.1 Decommissioning Objectives ........................................................................... 4-1 4.2 Decommissioning Activities .............................................................................. 4-1 4.3 Removal of Hazardous Materials ..................................................................... 4-3 4.4 Remediation Actions .......................... .... .... .... .... ... .... .... .... .... .... .... .... .... .... ....... 4-3 4.5 Decommissioning Work Controls ..................................................................... 4-3 4.6 Decommissioning Organization & Responsibilities ..... .... .... .... .... .... .... .... .... ...... 4-5 5.0 FINAL STATUS SURVEY ............................................................................................ 5-1 6.0 COMPLIANCE WITH THE RADIOLOGICAL CRITERIA FOR LICENSE TERMINATION

.................................................................................................................................... ~1 7.0 FINANCIAL ................................................................................................................. 7-1 8.0 ENVIRONMENTAL REPORT ... .... .... .... .... .... .... .... .... .... ... .... .... .... .... .... .... .... .... .... ........ 8-1 9.0 BIBLIOGRAPHY .......................................................................................................... 9-1 ATTACHMENT 1 - VBWR ENVIRONMENTAL REPORT ........................................................ 9-1

TABLE OF FIGURES/TABLES SECTION PAGE Figure 1-1: Regional Area Map ............................................................................................................. 1-6 Figure 1-2: VNC Area Photo ....... .... .. ... ... ... ..... .. ...... .. ..... ... ...... .. ..... ... ...... .. ..... ... ...... .. ..... ... ...... .. ..... ... ..... 1-6 Figure 1-3: VNC Site Area ..................................................................................................................... 1-7 Figure 1-4: VNC 300 Area ..................................................................................................................... 1-8 Figure 1-5: Topography Contour of Vallecitos Nuclear ..................................................................... 1-9 Figure 2-1: Application of the MARSSIM Process for the VBWR Characterization and LTP .... 2-2 Figure 2-2: VBWR Main Floor (Entry Level) ....................................................................................... 2-4 Figure 2-3: VBWR Basement Area ................................. .. .... .. ........ .. .... .. ........ .. .... .. ........ .. .... .. ........ .. .... 2-5 Figure 2-4: VBWR Outside Containment Areas ................................................................................. 2-6 Figure 2-5: VBWR Reactor Vessel Exposure Rates ........ ... ...... ... .... ... ...... ... .... ... ...... ... .... ... .............. 2-7 Figure 2-6: VNC Well Monitoring Locations ................................... ...... .. ...... .. ...... .. ...... .. ...... .. ........... 2-10 Figure 2-7: Ludlum 43-93 Beta Detection Calibration Curve ............ .. ...... ... .... ... ...... ... .... ... ............ 2-19 Figure 2-8: Ludlum 43-93 Alpha Detection Calibration Curve ........................................................ 2-20 Table 2-1: 300 Area Soil Sample Analytical Results ......................................................................... 2-8 Table 2-2: Monitoring Well Results of MW9S and MW9D (through February 2023) .................. 2-11 Table 2-3: Background Well MW10R Results (February 2023) ..................................................... 2-11 Table 2-4: Concrete Core Sampling Locations within Containment... ..... .. ...... .. ...... .. ...... .. .... ... ..... 2-12 Table 2-5: Radiochemical Analysis of Selected Cores .................................................................... 2-12 Table 2-6: HSA Identified Events Causing Contamination at VBWR. ........................................... 2-14 Table 2-7: Lead and Asbestos Abatement. .......................................................................................2-15 Table 2-8: Designation of Survey Areas for Characterizing VBWR .................. ...... .. ...... .. .... ... ..... 2-15 Table 2-9: VBWR Radionuclides of Concern and Fractional Abundances .................................. 2-16 Table 2-10: Characteristics of Selected Radiation Detection lnstruments ................................... 2-18 Table 2-11: Site-Specific Ludlum 43-93 Total Efficiency for Beta Emitters .................................. 2-19 Table 2-12: Site-Specific Ludlum 43-93 Total Efficiency for Alpha Emitters .......... .. ...... .. .... ... ..... 2-20 Table 2-13: Reference Material Background Determinations ..................................... ...... .. ........... 2-21 Table 2-14: VBWR Radioactivity Inventory to be Transferred ........ ... ...... ... .... ... ...... ... .... ... ............ 2-23 Table 3-1: Major Remaining Activities and Completion Dates ......................................................... 3-3 Table 3-2: Estimated Solid Waste Activity & Volume ........................ .. ...... ... .... ... ...... ... .... ... ...... ... .... .3-5 Table 4-1: ALARA Dose Estimate for Reactor Vessel Removai. ...... .. ...... .. ...... .. ...... .. ...... .. ...... .. .... A-4

LIST OF ACRONYMS AEC Atomic Energy Commission A LARA As Low As Reasonably Achievable BWR Boiling Water Reactor CFR Code of Federal Regulations CRD Control Rod Drive D&D Dismantlement and Decontamination DAW Dry Activated Waste DCGL Derived Concentration Guideline Level DPC Design package and checklist DPR Developmental Power Reactor dpm/100 cm 2 Nuclear disintegrations per minute per 100 square centimeters DQO Data Quality Objectives EPD Electronic Personal (Radiation) Dosimeter ESADA Empire State Atomic Development Associates Incorporated EVESR ESADA Vallecitos Experimental Superheat Reactor FSAR Final Safety Analysis Report FSS Final Status Survey ft3 cubic foot GE General Electric GEH GE-Hitachi Nuclear Group Americas, LLC GETR General Electric Test Reactor HEPA High Efficiency Particulate Air HRA High Radiation Area HSA Historical Site Assessment LPSDAR Limited Post Shutdown Decommissioning Activities Report LTP License Termination Plan MARSSIM Multi-Agency Radiation Survey and Site Investigation Manual mR/h MilliRoentgen per hour mrem/h Millirem per hour N/D Not Detectable (regarding presence of radioactive material)

NRC Nuclear Regulatory Commission NTR Nuclear Test Reactor pCi picocurie (1/1 ,000,000,000,000 (one trillionth) of a curie)

PPE Personal Protective Equipment PSDAR Post- Shutdown Decommissioning Activities Report QA Quality Assurance QAP Quality Assurance Plan RP Radiation Protection RWP Radiation Work Permit SNM Special Nuclear Material TEDE Total Effective Dose Equivalent TSA Total Surface Activity VBWR Vallecitos Boiling Water Reactor VNC Vallecitos Nuclear Center VSP Visual Sampling Plan; a software package for designing surveys developed by Department of Energy WEP Waste Evaporator Plant

LIST OF DEFINITIONS A LARA "as low as reasonably achievable," which means making every reasonable effort to maintain exposures to radiation as far below the dose limits as is practical.

BWR Boiling Water Reactor: a type of light-water-cooled nuclear reactor where the water is converted to steam within the reactor vessel, transported to the turbine generator for electricity production, and then recycled back into water by a condenser, to be used again in the heat process.

Characterization A type of survey that includes facility or site sampling, monitoring, and Survey analysis activities to determine the extent and nature of contamination. Characterization surveys provide the basis for acquiring necessary technical information to develop, analyze, and select appropriate cleanup techniques.

Data Quality Qualitative and quantitative statements derived from the DQO Objective (DQO) process that clarify technical and quality objectives, define the appropriate type of data, and specify tolerable levels of potential decision errors that will be used as the basis for establishing the quality and quantity of data needed to support decisions.

DAW Dry Activated Waste: Radwaste that is typically paper, wood, plastic, trash, air filters, metal, soil, concrete, asphalt, and used plant components, which without processing, contains essentially no free liquid.

Decommissioning The major steps that make up the reactor decommissioning process Process are: certification to the NRC of permanent cessation of operations and removal of fuel; submittal and implementation of the post-shutdown decommissioning activities report (PSDAR); submittal of the license termination plan (LTP); implementation of the LTP; and completion of decommissioning.

Derived The Derived Concentration Guideline Level (DCGL) is a Concentration radionuclide-specific surface or volume residual radioactivity level that Guideline Levels is related to a concentration or dose- or risk-based criterion. For the (DCGLs) NRC, DCGLs are the radionuclide specific activity concentrations that correspond to the release criterion (25 mrem/y) within a survey unit.

DQO Qualitative and quantitative statements derived from the DQO process that clarify study technical and quality objectives, define the appropriate type of data, and specify tolerable levels of potential decision errors that will be used as the basis for establishing the quality and quantity of data needed to support decisions.

Final Status Survey Measurements and sampling to describe the radiological conditions of a site, following completion of decontamination activities (if any) in preparation for release.

Historical Site The identification of potential, likely, or known sources of radioactive Assessment (HSA) material and radioactive contamination based on existing or derived information for the purpose of classifying a facility or site, or parts thereof, as impacted or non-impacted.

Investigation level A derived media- specific, radionuclide - specific concentration or activity level of radioactivity that: 1) is based on the release criterion, and 2) triggers a response, such as further investigation or cleanup, if exceeded.

MARSSIM The Multi- Agency Radiation Site Survey and Investigation Manual (NUREG-1575) is a multi-agency consensus manual that provides information on planning, conducting, evaluating, and documenting building surface and surface soil final status radiological surveys for demonstrating compliance with dose or risk-based regulations or standards.

Radioactivity The mean number of nuclear transformations occurring in a given quantity of radioactive material per unit time. The International System (SI) unit of radioactivity is the becquerel (Bq), denoting 1 disintigration per second. The customary unit is the curie (Ci), which is defined as 1 Ci =3.7 x 10 10 disintegrations per second.

SAFSTOR The alternative in which the nuclear facility is placed and maintained in a condition that allows the nuclear facility to be safely stored and subsequently decontaminated (deferred decontamination) to levels that permit release for unrestricted use.

Scoping Survey An initial survey performed to evaluate: 1) radionuclide contaminants,

2) relative radionuclide ratios, and 3) general levels and extent of contamination.

Total Effective Dose The sum of the effective dose equivalent (for external exposure) and Equivalent (TEDE) the committed effective dose equivalent (for internal exposure-).

8 HITACHI 1-1

1.0 INTRODUCTION

AND PURPOSE Pursuant to 10 CFR 50.82(a)(9), GE- Hitachi Nuclear Energy Americas LLC (GEH) has prepared this license termination plan (LTP) in support of the termination of GEH license DPR-1 for the Vallecitos Boiling Water Reactor (VBWR), located at the Vallecitos Nuclear Center (VNC), in Sunol, California. Formatting and content are consistent with Regulatory Guide 1.179, Standard Format and Content of License Termination Plans for Nuclear Power Reactors [1].

Because the VBWR has no Final Safety Analysis Report (FSAR), this LTP is being submitted as an "equivalent" document pursuant to 10 CFR 50.82(a)(9)(i). Additionally, as discussed in the GEH Vallecitos Nuclear Center Decommissioning Limited Post Shutdown Decommissioning Report (ADAMS Accession No. ML22264A325, ML22264A326, ML22264A327, ML22264A328),

September 21, 2022, this LTP serves as the license basis document for performing activities under§ 50.59 in lieu of an FSAR and therefore includes an updated accounting of decommissioning dismantlement activities performed pursuant to 10 CFR 50.59.

According to the Nuclear Regulatory Commission (NRC) letter, GE Hitachi Nuclear Energy, Vallecitos Nuclear Center- Response to the Description for Decommissioning the Shutdown Reactors (ADAMS Accession No. ML22066A569), April 8, 2022, termination of the VBWR license is contingent on the transferal of the remaining facility and its in-situ residual contamination to the authority of the GEH DR-1 0 license for the Empire State Atomic Development Associates Incorporated [ESADA] Vallecitos Experimental Superheat Reactor (EVESR) and approval of a 10 CFR 20.12 exemption to 10 CFR 50.82(a)(11)(ii) that will allow the VBWR DPR-1 license to be terminated without demonstrating that the facility and site have met the criteria for decommissioning in 10 CFR 20, subpart E.

This LTP specifically addresses the removal of the VBWR reactor vessel as the remaining component that is inseparable from the DPR-1 license. Removal of the vessel is planned to be performed pursuant to 10 CFR 50.59 in 2023. With the transfer of residual radioactive material from the VBWR license to EVESR license, there is no final status survey. Surveys are sufficient for quantifying the residual levels, as well as for determining radiological hazards and for establishing appropriate radiological controls for continued compliance with the Radiation Safety Standards of 10 CFR 20.

Once the DPR-1 license is terminated, the VBWR facility and all in-situ residual contamination under the authority of GEH DR-10 will be included in the EVESR LTP so that the final decommissioning of the VBWR will be coincident with the final decommissioning of the EVESR facility and the termination of the EVESR DR-10 license. This allows for the decommissioning of the VBWR, EVESR, and GE Test Reactor (GETR) facilities by the EVESR final license termination date of April 15, 2030.

1.1 Site Licensing The VBWR (DPR-1 Docket 50-18) and EVESR (DR-10 Docket 50-183) are both licensed as power reactors under Part 50, "Domestic Licensing of Production and Utilization Facilities," of Title 10 of the Code of Federal Regulations (CFR). Both facilities have permanently ceased operation, are in SAFSTOR, and all nuclear fuel has been removed from both reactor vessel's and from the VNC site.

The VBWR and EVESR and their separate reactor buildings are co-located within an approximate 3.2-acre area within the larger VNC site referred to as the "300 Area". The EVESR

8 HITACHI 1-2 DR-10 license was initially issued on November 13, 1961, and by design, EVESR used the secondary plant and many of the support systems and structures originally built for the VBWR.

General Electric (GE) was issued license CPPR-3 to construct and operate the VBWR on May 14, 1956, as Developmental Power Reactor (DPR). License CX-2 to operate DPR was issued July 29, 1957, and initial criticality was achieved on August 3, 1957. License DPR-1 was issued to GE on August 31, 1957, and full power was attained on October 19, 1957. The VBWR ceased operations on December 9, 1963, was defueled on December 24 of that same year. GE was issued a license to possess but not operate the VBWR reactor on September 9, 1965. The Atomic Energy Commission (AEC) issued GE an Order to Dismantle the VBWR on July 25, 1966. On October 22, 2007, license DPR-1 amendment 21 was issued, transitioning ownership of the VBWR from GE to GEH.

VNC DPR-1 License Condition 3a stipulates that the facility is to be possessed under the 10 CFR Part 50 portion of the license in the condition described in the Final Report on Deactivation of Vallecitos Boiling Water Reactor, dated February 5, 1965, (hereafter referred to as the Deactivation Report) [2].

The Deactivation Report [2] includes details of certain equipment and facilities that were at that time transferred to the authority of the EVESR license. The VBWR license retained the following:

• Reactor enclosure and contents.

• Stack (since demolished).

• Valve pit between the turbine and reactor buildings (pit located adjacent the VBWR containment with piping and components having been removed).

• Some wastewater and makeup water equipment and air lines (since removed for disposal).

• The reactor feed pumps and instrument air compressor located in the turbine building (since removed for disposal).

The VBWR afore-listed components not within the reactor building have been dismantled and removed (pursuant to 10 CFR 50.59 as applicable) consequent to decommissioning. Except for the VBWR ventilation system abutting the VBWR containment, no separate structures external to the reactor building remain under the authority of the license.

The VBWR facility, for purposes of this request, is therefore defined as the reactor enclosure, its contents, and remaining VBWR-specific plant components adjacent the reactor enclosure as stipulated in the Deactivation Report [2], notwithstanding the absence of these components. In summary, all residual radioactivity, other than that contained within the VBWR reactor containment, VBWR valve pit, and the VBWR ventilation system, whether from VBWR or EVESR operations, are under the EVESR license.

1.2 Transfer of VBWR Residual Contamination to EVESR Pursuant to 10 CFR 50.12(a), the Commission may, upon application by an interested person or upon its own initiative, grant exemptions from the requirements of 10 CFR 50 when the exemptions are authorized by law, will not present an undue risk to the public health or safety, are consistent with the common defense and security, and when special circumstances are present. Special circumstances are present, according to 10 CFR 50.12(a)(2)(ii), whenever application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule.

8 HITACHI 1-3 The underlying purpose of 10 CFR 50.82(a)(11) is to describe the requirements that must (usually) be met for license termination. §50.82(a)(11)(i) relates to dismantlement of the facility and therefore to license condition 3a of the VBWR license (the Part 50 portion of the license).

§50.82(a)(11)(ii) relates to decontamination of the facility, and license condition 3b (the Part 30 portion of the license). §50.82(a)(11)(ii) specifies that the results of the terminal radiation survey and other documentation show that the facility and site meet the requirements for release according to 10 CFR Part 20, subpart E.

For VBWR license termination, following the reactor vessel removal, residual radioactive materials will remain under the EVESR license until the EVESR license is terminated. The VNC site will not be released even until after termination of the EVESR license because of continued operations under NRC site Special Nuclear Material (SNM) license SNM-960 and California byproduct material license CA 0017-01. Therefore, application of the rule (1 0 CFR 50.82(a)(11)) is not necessary to terminate the VBWR license because the EVESR license will remain in effect and can accommodate both the Part 50 portion of the license (license condition 3a) and the Part 30 portion (license condition 3b). Since the residual materials will remain under an NRC license (meeting the decommissioning rule standard that the NRC complete decommissioning) and under the same conditions of use, the granting of an exemption from 10 CFR 50.82(a)(11) will be in accord with public health and safety and the common defense and security, as well as the public interest, and there will be no environmental impact. Therefore, transferring the remaining VBWR facility and residual contamination in the facility to the EVESR license renders 10 CFR 50.82(a)(11) unnecessary to achieve its underlying purpose.

All such materials in both the VBWR and EVESR Facilities are limited by license condition to byproduct materials contained in components and parts that comprise the facility. Therefore, transferring licensing authority of this material will not exceed any EVESR DR-1 0 possession limit. The VNC is to maintain authority of this material and will continue to do so as long as the material remains under the applicable NRC license at the VNC.

1.3 Site Location and Description The VNC site is located at 6705 Vallecitos Road (north side of State Route 84) in Sunol, Alameda County, California (see Figure 1-1). The site is east of San Francisco Bay, approximately 35 air miles east-southeast of San Francisco and 20 air miles north of San Jose.

The valley is at an elevation of 400 to 500ft (120 to 150 km) above sea level and is surrounded by barren mountains and rolling hills. The properties surrounding the site are primarily used for agriculture and cattle raising, with some residences, which are mostly to the west of the property. The nearest cities are Pleasanton, with a population of approximately 80,000, located to the north-northwest and Livermore with a population of approximately 90,000, located to the northeast. (Refer to Figure 1-1.)

The original VNC site is understood as an approximate 1600-acre site, which includes both the GE property (approximately 610 acres) and the VNC property (approximately 997 acres). The fenced area, commonly referred to as the Site Developed Area, encompasses the buildings and structures associated with all operations at the Site. The Site (Figure 1-2, Figure 1-3, and Figure 1-4) is bounded on the west, north, and east by hilly terrain; in some places, the hills are about 400ft (120m) above the general Site elevation. Vallecitos Road (SR 84) is at the southern boundary of the Site, from which an expanse of gently rolling grassland extends north for about 1.2 mi (2 km) at which point mountains form a northern barrier; completing the geographical encirclement of the Site.

8 HITACHI 1-4 The VBWR and EVESR licensed area, also known as the 300 Area, is approximately 3.2 acres within the Site Developed Area. Maps of the areas are included as Figure 1-2 through Figure 1-4.

In addition to the three shutdown reactor facilities within the VNC Site Developed Area, the following list describes other facilities at VNC. Locations for these facilities are shown in Figure 1-3.

• Building 102 contains the Materials Operations Laboratory (formerly the Radioactive Materials Laboratory), used for post-irradiation studies and research and development activities, and administrative offices.

• Buildings 102A and 102B contain facility equipment (ventilation, etc.) and offices for support to the site.

• Building 103 houses analytical chemistry laboratories and administrative offices.

• Building 104 contains Site storage and warehouse facilities.

• Building 105 contains offices and laboratories and houses the Nuclear Test Reactor (NTR) and another shielded cell that was formerly used as a critical experiment facility.

• Building 106 contains machine, sheet metal and facilities maintenance shops.

• Building 107 is the Hazardous Material Storage Building.

• Building 400 and 401 are currently leased to ManTech.

Temporary storage of solid radioactive materials and low-level radioactive waste is accommodated at the Hillside Storage Facility. Facilities are available on the Site for handling, sorting, and processing liquid and solid radioactive wastes generated at all VNC nuclear facilities. A liquid waste evaporator facility is located within the 300 Area; operation of the evaporator is covered by California byproduct material license CA 0017-01. A nonradioactive liquid waste treatment plant and sewage treatment plant are located in the southwest corner of the Site.

1.4 Overview of LTP Sections 10 CFR 50.82(a)(1 0) calls for an LTP that demonstrates: " ... that the remainder of decommissioning activities will be performed in accordance with the regulations in this chapter, will not be inimical to the common defense and security or to the health and safety of the public, and will not have a significant effect on the quality of the environment .... " To this end, the objective of this LTP is to outline the programs and processes for performing the identified decommissioning activities supporting the transition of the remaining VBWR reactor licensed facility to the EVESR reactor license to enable termination of the VBWR license.

The following summarizes the information contained in this LTP:

• Section 2, Characterization, provides a radiological characterization of VBWR, documenting the residual levels to remain, following reactor vessel removal. An estimate of the total residual radioactive materials levels is presented, supporting the transfer to the EVESR license.

• Section 3, Remaining D&D Activities, presents an overview of the remaining site dismantling and decommissioning activities. This section, supplemented by Section 4, identifies the major dismantlement activities that remains to be completed. The primary decommissioning activity to be completed is the reactor vessel removal and disposal.

The method by which the reactor vessel is to be removed is still to be determined. In brief, two options have been considered. The chosen option involves removal of the

8 HITACHI 1-5 intact vessel, which would require cutting an opening in the reactor building structure.

As indicated in the VNC Limited Post Shutdown Decommissioning Activities Report (LPSDAR), GEH will review this activity pursuant to §50.59. Additionally, this section includes an estimate of associated occupational radiation dose for the reactor vessel removal and an overall projected volume of radioactive waste.

• Section 4, Site Remediation Plan, identifies different decontamination techniques that may be used; it describes the activities for removal of lead and asbestos contaminated materials. The decommissioning activities will be performed under the direction and oversight of GEH and VNC site management and programs. The VNC organization and programs supporting the decommissioning are discussed, including the VNC Change Authorization Process for work planning and control, the Radiation Protection (RP) Program, and the Quality Assurance (QA) Program.

• Section 5, Final Status Survey, acknowledges that with the transfer of all residual radioactive materials to EVESR, the final status survey for the 300 Area will be performed during the decommissioning and license termination of EVESR.

• Section 6, Compliance with the Radiological Criteria for License Termination, demonstrates that residual levels of radioactivity for VBWR are minimal and do not pose a radiological safety and health concern pending a final site remediation and license termination under EVESR.

• Section 7, Financial, provides an update to the estimated cost for decommissioning VBWR.

• Section 8, Environmental Report, demonstrates that the decommissioning activities will not pose an unacceptable environmental impact and identifies measures that will be taken to ensure compliance with California environmental regulations.

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8 HITACHI 1-8 0 100 FIGURE 1-4: VNC 300 AREA

8 HITACHI 1-9 FIGURE 1-5: TOPOGRAPHY CONTOUR OF VALLECITOS NUCLEAR

8 HITACHI 2-1 2.0 CHARACTERIZATION 2.1 Site Characterization 2.1.1 Approach The characterization of VBWR has generally followed the Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM) -described process (4). This process included a Historical Site Analysis (HSA) and Scoping surveys, followed by Characterization surveys.

Characterization will continue during the dismantling and remediation process, with surveys post vessel removal being used for a final radiological characterization prior to VBWR license termination request. Figure 2-1 depicts the application of the MARSSIM process as applied for VBWR license termination, where residual radioactivity is to be transferred to EVESR.

MARSSIM final status surveys (FSSs) will be performed during the EVESR license termination.

Therefore, the radiological characterization surveys to support the license transfer are represented by those summarized below, which have been based on the HSA, and Characterization surveys. [3] [4] [5]

The removal and disposal of the reactor vessel, with associated internals, will leave the VBWR facility with minimal residual radioactive materials under License DPR-1, Docket 50-18. As identified above, the general MARSSIM approach was used to identify and characterize the remaining radioactive materials. The following sections describe the process, followed by a quantification of the remaining radioactive materials for transfer to the EVESR license.

8 HITACHI 2-2 Historical Site Assessment (HSA)

Seeping/Characterization Surveys license Termination Plan (LTP)

Compliance with Radiological Criteria (QAPP, DQOs, MQOs, & Survey Plan)

Update Remedial Action ALARA Characterization Support Surveys Data Quantifying Residual Radioactive Materials Levels for Transfer to EVESR License Termination FIGURE 2-1: APPLICATION OF THE MARSSIM PROCESS FOR THE VBWR CHARACTERIZATION AND l TP The general criteria used to classify the identified impacted or non-impacted areas was drawn from the regulatory guidance of MARSSIM as follows [6]. The HSA collected, organized, and evaluated information that described the VBWR in terms of physical configuration and the extent to which the reactor site was radioactively contaminated as a result of plant operations and decommissioning activities. All VBWR structures associated with the site were considered impacted. Radiological seeping and characterizations surveys were performed. Non-radiological hazards were evaluated. The following addresses these processes and results.

2.1.2 Boundaries of the VBWR As identified, the DPR-1 license is a license to possess, but not operate, the VBWR facility. License Condition 3a stipulates that the facility is to be possessed under the 10 CFR Part 50 portion of the license in the condition described in the Deactivation Report [2]. The VBWR licensed facility is therefore defined as the reactor enclosure and contents and all plant components outside of the reactor enclosure as stipulated in the Deactivation Report [2]. All reactor support systems inside the containment structure have been removed except for the reactor vessel; the main floor and the basement areas of the VBWR are depicted in Figure 2-2 and Figure 2-3, also reflecting the ambient direct exposure rates as measured during characterization. The HEPA filter ventilation system, located external to the containment, is to

8 HITACHI 2-3 remain in-place and will be remediated during the EVESR license termination. Additionally, the valve pit will remain. The VBWR stack has been removed, accompanied by excavation and removal of associated ducting along the northeast area outside of containment. These components of the plant facility no longer exist; see Figure 2-4.

Figure 1-2 shows the boundary of the site and plant exclusion area and Figure 2-2, Figure 2-3, and Figure 2-4 identify major components within and outside the reactor enclosure. Remaining areas for decontamination or remediation are those with elevated exposure rates, specifically for the bioshield following removal of the reactor pressure vessel. Need for remediation will be evaluated considering ALARA for potential worker exposures and minimization of the potential migration/dispersion pending final remediation during EVESR license termination. General area radiation exposure rates are shown in Figure 2-2, Figure 2-3, and Figure 2-4.

A characterization of the reactor vessel was performed by OW James Consulting [7]; the results of this characterization were used for estimating exposure rates associated with the reactor vessel. A sketch for the reactor vessel with estimates of dose rates is shown as Figure 2-5.

This characterization will be used for planning the removal, packaging, transportation, and disposal; however, since the vessel will have been removed prior to license termination, its radionuclide inventory is not included in the characterization of the residual radioactivity, as presented in Section 2.3.

8 HITACHI 2-4 SPENT FUEL POOL VBWR MAIN FLOOR E>

LEGEND: All measurements in mR/h; survey source indicated below.

0 RSR2022-0l-0072, May 26, 2022

(,..-o-.0-8--0--'.=

lS""'J RSR2022-01-0074, May 31, 2022 FIGURE 2-2: VBWR MAIN FLOOR (ENTRY LEVEL)

8 HITACHI 2-5 VBWR BASEMENT AREA 0

LEGEND: All measurements in mR/h; survey source indicated below.

0 RSR2022-01-002, and 0181, May 26, 2022 WALL - . Scoping 035, Aug. 15, 2022 FIGURE 2-3: VBWR BASEMENT AREA

8 HITACHI 2-6 exists in this area, ~

rest is excavated.

HEPA Filter System Valve Pit VBWR CONTAINMENT BUILDING

./

Personnel

., )

~- ""'

LEGEND: All measurements in mR/h; survey source indicated below.

0 .4 RSR2022-01-0032, May 3, 2022 FIGURE 2-4: VBWR OUTSIDE CONTAINMENT AREAS

8 HITACHI 2-7 DWJ Adjusted Locatio n Exposure Rate (mRjhr)

(

14'- Contact 4.93E+01 ' ~ ~~

16'- Contact 5.18E+Ol I.~ _ -..:

1a*- Contact 2.48E+Ol ~

19.75'- Contact 2.54E-Ol f- 5 ' ~- ,

6'- 1a" (General Area) 3.29E-Ol a*- ta* (General Area) 1.03Et00 10' -1a* (General Area) 3.58E+OO 12'- 1a* (General Area) 1.06E+Ol 14' -1a" (General Area) 1.69E+Ol 16'- 18" (General Area) 1.68E+Ol 10 '

ta*- 18" (General Area) 1.03E+Ol 19.75'- 18" (General Area) 2.30E+OO 6'- 2 Meters 9. 71E-Ol a*- 2 Meters 1.74Et00 10'- 2 Meters 2.86E+OO 1-15' 12'- 2 Meters 4.09E+OO 14' - 2 Meters 5.07E+OO

'\,.

16'- 2 Meters 4.86Et00 1a*- 2 Meters 4.00Et00 19.75' - 2 Meters 2.78Et00 Bottom- Contact 7.97E-Ol Bottom- 18" 2.26E-01 Bottom- 2 Meters 6.82E-02 CTO Source: DWJamesCRPR-0034, March 22, 2022 FIGURE 2-5: VBWR REACTOR VESSEL EXPOSURE RATES 2.1.3 Area 300 Surface Soil and Groundwater Samples While the outside environs for the 300 Area, except for the VBWR sump and HVAC system and ducting, are included under the EVESR license. Additional characterization and, as needed, remediation will be addressed as part of the EVESR decommissioning and license termination process. An initial evaluation of radioactivity levels in the groundwater and surrounding ground surface for the 300 Area was conducted in 2019 [8].

Soil Samples Six soil samples were taken around VBWR between June 19-22, 2019. Two samples (VBWR

1. 1 and VBWR #3) were taken east and one (VBWR #2) taken west of the VBWR at assumed background locations. Three samples were taken within the VBWR area, including one (VBWR

1. 6) taken near the fence line where elevated external dose rates were measured. Cesium-137 was identified in all VBWR soil samples by gamma spectroscopy. VBWR #1 through VBWR #5

8 HITACHI 2-8 identified Cs-137 at levels below 1 pCilg, which is considered representative of background and with no apparent increase from VNC activities. Sample VBWR #6 identified elevated Cs-137 at 20.9 pCilg, which is considered above background and indicative of a VNC contribution. VBWR

1. 6 was further analyzed for hard to detect radionuclides listed in Table 2-1.

Six soil samples were taken around EVESR between June 19- 21, 2019. One sample (EVESR

1. 1) was taken east and one (EVESR #2) was west of the EVESR at assumed background locations. Four samples were taken within the EVESR area, including one (EVESR #5) taken near the equipment hatch and one (EVESR #6) taken near a tank pad. Cesium-137 was identified in all EVESR soil samples, based on gamma spectroscopy analysis. EVESR #1 through EVESR #5 identified Cs-137, at levels below 0.3 pCilg. EVESR #1 had a positive Co-60 at 4.41E-02 pCilg. EVESR #5 had a positive Co-60 at 3.64E-02 pCilg. The Cs-137 component was representative of background (fallout) radioactivity and not necessarily indicative of a Vallecitos activities. EVESR #6 identified elevated Cs-137 (9.4 pCilg) and Co-60 (4.81 E-02 pCilg). This sample indicated elevated activity attributable to VNC operations. This sample was further analyzed for hard to detect radionuclides with results identifying Ni-63.

A summary of the radiochemical analyses, as performed for VBWR #6 and EVESR #6 being the higher activity samples, is presented in Table 2-1, below. The high abundance of Ni-63, compared with Cs-137, will be further examined during the follow-up EVESR decommissioning, as well as presence for other hard-to-detect and TRUs.

TABLE 2-1: 300 AREA SOIL SAMPLE ANALYTICAL RESULTS VBWR#G EVESR#G Analyte Analytical 2a Analytical 2a Result Uncertainty Result (pCi/g) Uncertainty (pCi/g)

Cs-137 2.09E+01 3.26E-01 9.40E+OO 2.35E-01 Co-60 N/D N/D Am-241 N/D N/D Cm-243/244 N/D N/D Pu-238 N/D N/D Pu-239/240 N/D N/D Sr-90 N/D N/D C-14 N/D N/D Ni-63 1.30E+03 5.23E+02 6.77E+02 2.77E+02 TOTAL 1.32E+03 6.86E+02 N/D - not detected

8 HITACHI 2-9 Ground Water Samples Groundwater is monitored by collecting and analyzing samples from several wells located at VNC. The locations of wells are shown in Figure 2-6 (Figure 2-6 adapted from Bibliography Reference [9]). Well MW-9 was installed in 2019, consisting of nested monitoring wells MW-9S/9D, with depths of 45ft below grade surface (bgs) and 88ft bgs, respectively, with screen intervals of 25 to 45 ft bgs and 68 to 88 ft bgs. These nested wells are downgradient of the EVESR and VBWR facilities.

The design and installation of these two wells are described in a July 22, 2019, letter report from Brown and Caldwell to GE-Hitachi. [9] As described:

An assessment of groundwater flow across the Site was conducted in October 2018.

This assessment indicates that groundwater flows roughly southwest across the site at an average gradient of 0.03 feet/foot, as shown on Figure 3 of the Work Plan. The basement depths below grade for the three shutdown reactors are approximately 20 feet for the GETR reactor, 30 feet for the VBWR reactor, and 80 feet for the EVESR reactor.

For these reasons, the objective was to install one shallow monitoring well southwest of the GETR reactor (MW-8) and one shallow (MW-9S) and one deep (MW-90) nested set of monitoring wells southwest of the EVESR and VBWR reactors, which are located immediately adjacent to each other.

Initial sample samples were collected in 2019 and analyzed by gamma spectral analysis, gross beta, gross alpha and tritium (H-3) analyses. A follow-up analysis for Sr-90 was performed.

The gamma spectral analyses did not detect any radionuclides other than Bi-210 and Pb-210, which are naturally occurring radionuclides from the U-238 decay. Samples were positive for gross beta; levels were within the expected background variation with no indication of a plant-related component. Tritium, gross alpha and Sr-90 were below detection level.

MW-9S/9D was added to the VNC environmental monitoring program with quarterly samples collected for the first 2 years (2021 and 2022) with annual sampling continuing. Samples are analyzed for tritium, strontium-90, gross alpha, and gross beta. Detection of elevated levels would direct additional analysis by gamma spectroscopy and hard-to-detect radionuclides as determined needed. The results of well water samples from MW-9S and MW-90 are summarized in Table 2-2. A background well, MW-10R, was added in 2022 which is also sampled on an annual basis, see Figure 2-6 for location and Table 2-3 for results.

Samples to date have not indicated any contribution to groundwater radioactivity levels attributable to VBWR or EVESR. Groundwater monitoring is continuing and will be further evaluated during the EVESR decommissioning.

HITACHI 2-10

• -~

' -~ . . w.vm UHI!

lt2 '

• • * -oO 0o Legend

-$- Monitoring Well Network No Scale FIGURE 2-6: VNC WELL MONITORING LOCATIONS

8 HITACHI 2-11 TABLE 2-2: MONITORING WELL RESULTS OF MW95 AND MW9D (THROUGH FEBRUARY 2023)

Sample Data MW9S (4S-1E2P1) MW9D (4S-1 E2P2)

Gross Gross Sr-90 Tritium Gross Gross Sr-90 Tritium Alpha Beta (pCi/L) (pCi/L) Alpha Beta (pCi/L) (pCi/L)

(pCi/L) (pCi/L) (pCi/L) (pCi/L) lnitial2019 sample 2.23 2.96 -0.8 109 105 9.21 -0.3 310 February-21 1.85 1.62 0.51 200 1.21 1.05 0.65 286 May-21 4.45 1.52 0.21 631 1.39 1.80 0.05 7 August-21 4.90 2.46 0.65 453 4.77 3.13 0.81 108 November-21 0.98 3.64 0.13 161 2.77 3.35 0.16 204 February-22 3.95 2.5 0.63 726 4.51 2.92 0.67 362 May-22 4.90 2.59 1.87 566 4.63 3.04 1.10 397 August-22 2.82 4.71 0.51 308 4.81 2.22 1.07 68 November-22 3.96 0.16 0.76 363 2.73 1.68 0.48 11 February-23 0.86 1.02 0.00 243 2.33 3.58 0.84 514 Note: Analytical results are presented as the "as measured" values, where tritium minimum detection activity (MDA) ranged from a nominal 180 to 300 pCi/1 and Sr-90 MDA ranged from a nominal 0.2 to 1 pCi/1.

TABLE 2-3: BACKGROUND WELL MW1 OR RESULTS (FEBRUARY 2023)

Gross Sample Data Alpha B t Gross I Sr- 90 I Tn"f1um (pCi/L) (p~i/~) (pCi/L) (pCi/L)

I I Result 3.07 1.49 0.75 404

+/-2 Sigma 1.25 0.68 0.12 86 Error MDA 0.57 1.54 0.26 183 2.1.4 VBWR Containment Concrete Core Samples In April of 2021, fourteen (14) concrete core samples were taken from the VBWR in the containment basement locations and on the basement level, as listed in Table 2-4. The samples

8 HITACHI 2-12 were collected to a nominal 8-inch depth. These core samples were analyzed by the Vallecitos on-site radiochemistry laboratory for gamma emitters. Cs-137 was identified in all of the samples, and smaller quantities of Co-60 were identified in many (but not all) of the core samples. Three of the samples with highest results (VBWR Core C1-3A, C2-1A and C3-3A) were analyzed for full Part 61 radionuclide analysis and those results are presented in Table 2-5.

No radionuclides other than Co-60 and Cs-137 were identified. All hard-to-detect (Ni-63, alpha emitters or transuranic radionuclides were identified.

TABLE 2-4: CONCRETE CORE SAMPLING LOCATIONS WITHIN CONTAINMENT LSamf. e 1 I Number of I Description of Location oca 1on Cores Cylindrical and spherical C-1 5 exterior concrete walls Interior concrete walls C-2 3 supporting the reactor cavity C-3 3 Interior concrete columns C-4 3 Main floor concrete slab TABLE 2-5: RADIOCHEMICAL ANALYSIS OF SELECTED CORES vcs Sample GEL Lab ID INuclide I Activity (~Ci/g) and (%) of Total IU 2

• t nee amY I MDA (IJCi/g)

ID I 1.02E-03 5.81 E-VBWR Core C 1-21 Pl256-2 Cs-137 (99.8%) 2.28E- 4.60E-06 07 3A Co-60 06 (0.2%) 3.90E-07 1.31 E-07 3.28E-04 VBWR Core C2- 5.26E-07 21 Pl256-6 Cs-137 (99.7%) 3.98E-06 1A 3.60E-07 Co-60 1.00E-06 (0.3%) 4.39E-07 21 Pl256- VBWR Core C3- 1.49E-04 2.98E-07 11 3A Cs-137 (100%) 2.46E-06 2.2 Historical Site Assessment The HSA [5] for the shutdown reactors, which included VBWR, documented those events and circumstances that occurred during the operating history of the facility and could have contributed to the contamination of the site environs above background levels. Changes in the radiological status of the site that may occur after the HSA are evaluated as part of the

8 HITACHI 2-13 decommissioning activities, any events, leading to significant changes in radiological conditions, will be further evaluated with an update to the characterization during the license termination process. Due to the advanced state of decommissioning, the remaining decommissioning activities described in Chapter 3 and 4 are not expected to significantly alter the characterization, except for the removal of the reactor vessel, which represents the majority of the currently remaining radioactive material.

2.2.1 Personnel Interviews Personal interviews of current and former site personnel were held during the site inspection and via telephone during the HSA process. Personnel were selected based on their employment history at the site. Personnel were interviewed that held positions in maintenance, qualified reactor operators, and radiation protection. Undocumented events were not discovered during this process, but the interviews did prove helpful in assessing the historical operations. In total eleven current and former employees were contacted and seven were interviewed.

2.2.2 History of Past Operations and Unplanned Events The HSA reviewed available historical documents that provided information on the potential releases of radioactive material to the environment. Numerous documents discuss tasks that raise the potential for an unplanned release of radioactive material. Table 2-6 summarizes the identified events during past VBWR operations that could have resulted in facility and/or surrounding area contamination. The HSA did not identify any new areas of potential residual contamination that were not considered within the bounds of the 300 Area as being evaluated for license termination.

For past events causing potential contamination outside of the 300 area, characterization would be required as part of the EVESR decommissioning. For example, Table 2-6 identifies an event where water was observed overflowing from the thousand-gallon VBWR main waste sump to a sewer manhole, which discharged into the artificial reservoir via a tile line and open ditch. Any residual radioactive materials from this event, as well as any future events as may occur during VBWR and EVESR decommissioning, will be characterized for the EVESR decommissioning.

Such characterization and activity are to be included in the EVESR LTP.

HITACHI 2-14 TABLE 2-6: HSA IDENTIFIED EVENTS CAUSING CONTAMINATION AT VBWR Location Date of Type of Incident Incident Description Incident Incident Summary Environment 11/11/1957 Facility and Radiation A Health Physicist, working near the BWR Environment Incident No.3 enclosure, observed expulsion of steam and Contamination gallon quantities of water from the BWR air exhaust stack. The water fell on and ran off the turbine building roof. A substantial quantity of water was observed flowing over the ground in the direction of the pond.

Reservoir 1/18/1958 Facility and Radiation Water was observed overflowing from the Environment Incident No. 4 thousand-gallon VBWR main waste sump to a Contamination sewer manhole, which discharged into the artificial reservoir via a tile line and open ditch.

Soft water flow to the waste sump resulted in an estimated discharge of 6,000 gallons of a mixture of contamination and soft water.

Reactor 2/1/1958 Facility Radiation VBWR staff members and a Health Physicist Enclosure Contamination Incident No. 5 entered the reactor enclosure for the purpose of area clean-up and removal of filter cartridges from the No. 1 clean-up filter. Surveys of the main floor area indicated the presence of high-level particulate contamination in areas adjacent to the reactor missile shield.

Reactor 9/4/1958 Personnel Radiation An electronics technician and two VBWR staff Enclosure Contamination Incident No. 7 members entered the reactor enclosure basement to replace faulty cable on the safety amplifier. Reactor water was found dripping from leaks in the sight glass level system located above the immediate work area.

Reservoir 11/10/1958 Facility and Radiation Water was observed overflowing from the VBWR Environment Incident No. 8 main waste sump to a sewer manhole, which Contamination discharged into the artificial reservoir via a tile line and open ditch. An estimated 1200 gallons of a mixture of contaminated sump water and reactor make-up water was discharged to the reservoir.

Environment 2/10/1959 Personnel, VBWR Fuel A fuel element in the VBWR ruptured, releasing Environment, Element fission products to the reactor steam system.

and Facility Rupture The resulting radioactive isotope concentration Contamination in the off-gasses activated the off-gas stack monitor and the reactor scrammed automatically.

Very little contamination was found on the vegetation around the reactor. Also, radiation levels spiked within the facility.

Reactor 4/27/1959 Personnel and Radiation Transfer of a defected fuel element, which had Enclosure Facility Incident No. been in the VBWR core, was scheduled from the Contamination 11 VBWR to the RML pool in Building 102 for visual examination. Spots of contamination were found in the airlock, on the concrete pad outside the airlock, on the forklift, and cask.

8 HITACHI 2-15 2.3 Characterization 2.3.1 Lead and Asbestos Hazards Assessment VBWR reactor building piping systems and equipment were removed during earlier remedial action. In support of planned reactor vessel removal, an industrial hygiene seeping survey was performed in 2022 [10]. Lead and asbestos abatement were completed subsequently within the VBWR reactor building. A Certification of Abatement Clearance was provided [11]. Table 2-7 summarizes results of the abatement.

TABLE 2-7: LEAD AND ASBESTOS ABATEMENT Material Location I Friable/ I Estimated I (Area) Nonfriable Quantity Asbestos-Containing/Contaminated Surrounding Concrete/insulating Blocks and VBWR Reactor Friable 30 cubic yards Associated Debris Vessel Surrounding Not Lead Bricks VBWR Reactor 2 cubic yards Applicable Vessel Surrounding All horizontal Lead and Asbestos Dust VBWR Reactor Friable surfaces in the Vessel basement 2.3.2 Radiological Characterization The areas and items, as identified in Table 2-8, are identified as those remaining for VBWR where characterization should be conducted to support the evaluations needed for the license termination. Generally, all open land areas within the 300 Area are considered part of the EVESR license; therefore, VBWR has no open land areas designated within its license.

TABLE 2-8: DESIGNATION OF SURVEY AREAS FOR CHARACTERIZING VBWR Area/Item Description Level 1 - Floor, Walls & Dome Basement- Floor, Walls, & Ceiling Basement Sump Personnel Portal Equipment Portal Valve Pit Fuel Pool Bioshield Embedded Piping HEPA System Stack Pad

8 HITACHI 2-16 2.3.2.1 Site Specific Radionuclides of Concern An important element of characterization is the identification of the radionuclides of concern.

Select smear samples collected during the Characterization Surveys [3] were analyzed for important radionuclides, including hard-to-detect radionuclides. A review of the radioanalytical results yielded the following radionuclides of concern (ROCs) in Table 2-9, along with estimated fractional abundances.

TABLE 2-9: VBWR RADIONUCLIDES OF CONCERN AND FRACTIONAL ABUNDANCES Radionuclide I 2022 Fractional Abundance H-3 N/D*

C-14 0.022 Ni-59 N/D Co-60 0.002 Ni-63 0.081 Sr-90 0.026 Nb-94 0.012 Tc-99 N/D Cs-137 0.793 Eu-152 0.047 Eu-154 N/D Eu-155 N/D Pu-238 0.008 Pu-239/240 0.003 Pu-241 N/D Am-241 0.006 Cm-243/244 N/D

• N/D- Not Detected 2.3.2.2 2022 Site Characterization Survey The objective of the 2022 radiological survey [3] was to assess the nature, degree, and extent of radiological contamination in the reactor containment (Building 301) using MARSSIM guidance.

The characterization activities were guided by "Site Characterization Plan," which used the MARSSIM Data Quality Objective (DQO) process to establish the necessary requirements and methods for obtaining high quality characterization data [12]. The scope of the Characterization Survey follows:

• Identify and quantify the nature and extent of radiological materials

• Determine the distribution of radioactive material contamination in each area that contained radioactive materials contamination

• Obtain data to provide guidance for waste management planning

• Provide information to support the development of site-specific Derived Concentration Guideline Levels (DCGLs)

8 HITACHI 2-17

• Provide a decision-making basis for transfer of residual contamination to the EVESR license.

2.3.2.3 Methodology The survey package development involved performing walk-downs of each area. During the walk-down, details regarding the physical survey area, such as type of area (structure, system, or environ), surfaces in the area (wall, floor, ceiling, concrete, metal, or other feature) and dimensions. Data from previous seeping surveys and evaluations were reviewed and used as appropriate. Each survey package contained the following seven sections of information:

1. Detailed description of the survey area.

2. Photographs, drawing, or drawings of the survey area.

3. Summary data from operational surveys or previous surveys.

4. Characterization survey instructions-types and number of survey measurements and/or samples prescribed for the survey.

5. Survey instrument requirements and any special tools.

6. Health and safety requirements

7. RWP requirements.

The packages generated for each survey unit are called Design Package and Checklists (DPC).

DPC's were created utilizing Visual Sample Plan (VSP) Software, which generates sample locations when building specifications are inputted. DCPs are maintained at VNC and available for review/inspection as needed.

2.3.2.4 Survey Instrumentation Survey instrumentations were selected that provided detection capability for the suite of radionuclides expected to be present, including gross alpha and beta/gamma detection for surface measurements and gamma detection for personnel dose estimates, outside surveys, and building area dose rate measurements. As identified in the Characterization Report [3], a summary of the survey instrumentation, radiations detected, detector type, calibration source and use as used during the Seeping and Characterization surveys is provided in Table 2-10.

8 HITACHI 2-18 TABLE 2-10: CHARACTERISTICS OF SELECTED RADIATION DETECTION INSTRUMENTS Instrument I Detector Mode I I Radiation Detected I Detector Type I Calibration Source I Use Ludlum 2360 Surface static/

43-93 (100 cm 2 Alpha-Beta ZnS(Ag) Th-230/Tc-99 scan area) measurements Nal Area exposure Ludlum 19 Gamma Cs-137 (IJR/hour) measurement Pu-239 Tc-99 Ludlum 3030E Removable Swipe Smear ZnS(Ag) 43-10-1 Alpha-Beta counting 1-131 C-14 Ludlum 2221 Area Exposure Gamma Nai(TI) Cs-137 44-10 Measurements Measurements with the Ludlum 43-93 1 , which was used exclusively for the static and scan measurements in the characterization survey, are recorded independent, i.e., measurements of alpha and beta emissions are discriminated and recorded separately. To obtain a weighted instrument efficiency, using guidance from NUREG-1507, Rev. 1, the radionuclide fractional abundance is multiplied by the detection efficiency for the radionuclide emissions and then by a surface efficiency [13]. Alpha and beta efficiencies for the ROCs were derived using data from NUREG-1507, Rev. 2, Calibration Curves for the Ludlum 43-93 detector. Figure 2-7 and Figure 2-8 present the beta and alpha detection curves with energies required by VBWR ROCs. Four beta/gamma emitting radionuclides; Co-60, Nb-94, Cs-137, and Eu-152; were added to the curve developed in NUREG-1507, illustrating the VBWR ROCs. For these radionuclides, quantifying the beta activity is an appropriate means of quantifying total activity as the gamma emission accompanies beta decay.

1 The Ludlum model43-93 is a dual phoswich ZnS(Ag)/plastic scintillator detector (1.2-mg/cm2 Mylar window)

8 HITACHI 2-19 Beta Detector Source Calibration Curve 0.00 +--'--=---+-----1---+----+-----1,__----l....+--___;'-+----l 0 200 400 600 800 1,000 1,200 1,400 1,600 Maximum Beta Energy (keV)

FIGURE 2-7: LUDLUM 43-93 BETA DETECTION CALIBRATION CURVE Source: NUREG-1507, Rev. 2, Figure A-13.

The site-specific weighted total efficiencies for beta/gamma emitters and alpha emitters are presented in Table 2-11 and Table 2-12, respectively, with specifics for each radionuclide.

TABLE 2-11: SITE-SPECIFIC LUDLUM 43-93 TOTAL EFFICIENCY FOR BETA EMITTERS Radionuclide I 2022 Fractional I Instrument Efficiency I Source Efficiency Iw~~9* ht d

)e Abundances (Ei) (Es) Tot C-14 0.022 0.15 0.25 0.0008 Co-60 0.002 0.26 0.25 0.0001 Ni-63 0.083 0 0.25 0.0000 Sr-90 0.027 0.38 0.5 0.0050 Nb-94 0.012 0.31 0.5 0.0019 Cs-137 0.807 0.31 0.5 0.1251 Eu-152 0.048 0.38 0.5 0.0091 Sum 1.00E+OO Total Efficiency 0.142 For the VBWR alpha-emitting ROC, the alpha energy for Pu-238 did not fall within the original curve. Its efficiency was extrapolated at the apparent asymptote at the same level for Am-241.

8 HITACHI 2-20 Alpha Detector Source Calibration Curve 0.50 ..

0.45 II l 'fh-231 ~663-~V I I I Pu-239,Sl 9kevb m-241 5487 ke r:'\ l

' \ ~

1 I/

• I 1

0.40 I u -.,a-'4188" f V I I T I  :

0.35 f I I I _. ~

I il 1 Pu-238.5580 k v/ y I

~~ ~ Il I tI / II

"'o

,.:;* 30 u

go.2s

• u

,* f I ,- t f

j ! II

' 1 ! I I I l t

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I cmJ 43 5893 k J

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t- 1 I

!E 0.20 1 1 w

II fy= 7E108x 2

~-~~~Bx 1.7f 81 I I

-co.1s G)

E J I

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t l

o:z 11 I~ I I I rI It 1-20.10 ti

.5 0.05 I Ill f 11 I ,. J

f. I I 0.00 II I !II 'I 1* I l I I I II I II I I I 4,000 4,200 4,400 4,600 4,800 5,000 5,200 5,400 5,600 Alpha Energy (keV)

FIGURE 2-8: LUDLUM 43-93 ALPHA DETECTION CALIBRATION CURVE Source: NUREG-1507, Rev. 2, Figure A-14.

TABLE 2-12: SITE-SPECIFIC LUDLUM 43-93 TOTAL EFFICIENCY FOR ALPHA EMITTERS Frae f 1ona1 Instrument Source W e1. hte d Rad1onuchde I Ab 2022 d un ances (Alpha Only)

I . .

Eff1c1ency (Ei)

I . .

Eff1c1ency (Es)

I (9 )

Erot Pu-238 0.48 0.44 0.25 0.053 Pu-239/240 0.15 0.43 0.25 0.016 Am-241 0.37 0.44 0.25 0.041 Sum 1.00 Total Efficiency 0.110 2.3.2.5 Reference Material Typical background (dpm/100cm 2) values for concrete and metal surface material were obtained from locations and representative materials considered to be unimpacted. Ten locations were measured for both concrete and steel. These values are shown in Table 2-13 with application of the efficiencies derived above to establish the cpm values.

8 HITACHI 2-21 TABLE 2-13: REFERENCE MATERIAL BACKGROUND DETERMINATIONS 2.4 Estimate of VBWR Radiological Inventory for Transfer to EVESR This section details the method used to estimate the types and quantities of radioactive material associated with VNC VBWR that are to be transferred to the EVESR DR-10 License.

Data from the Scoping and Characterization Surveys, along with subsequent follow-up surveys conducted in 2023, were used to estimate the types and quantities of radioactive material remaining within the VNC VBWR licensed facility that are to be transferred to the EVESR DR-10 License.

The 2022 Characterization Report [3] provides an overview of radiological conditions. However, to determine the residual levels to support transfer to EVESR, the data from the individual Work Packages that documented the individual survey unit measurements were used. These measurements were evaluated using the instrument efficiencies, as determined above, to develop radionuclide-specific inventory. These Work Packages summarize the radiological condition of the facilities prior to removal of the pressure vessel. The current analysis is based on measurement data reflective of the day the measurements were taken. While additional radioactive decay from time of surveys to time of transfer will occur, this decay reduction is considered negligible considering the longer half-lives (> 4 years) for the ROCs. The Work Packages are available at the site for inspection and will be retained as part of the decommissioning records.

The Work Packages document the surface activity measured for the various survey areas.

Measured values provide the Total Surface Activity (TSA) for both alpha and beta/gamma activities (average values and percentages of total). The work packages included sketches that document the size of the areas measured. Table 2-14 provides a summary of the calculated total beta/gamma and alpha surface activity determined by multiplying the average activity (dpm/100 cm 2) times the area of the object/area.

There are multiple instances of penetrations and embedded piping; each was assigned a contamination level based on best estimates using relevant survey data from associated surfaces/areas:

• There are 10 embedded pipes from the containment building to the exterior Valve Pit:

all blanked but with an assumed average of 5" diameter and approximate length of 4' from the inside of containment to the inside of the Valve Pit. The total pipe area was 56 square feet. Note that three of them are labeled as spare. The contamination level in the pipes was assumed to be the same as in the valve pit.

• There are 10 penetrations near grade of the containment building ranging from 6" to 18" in diameter; the number of individual pipes (electrical conduit, air lines, etc.) is 135.

Each was assumed to be 0.5 feet long for a total area of 52 square feet. The contamination level on the main floor was assumed to be the same as within these pipes.

8 HITACHI 2-22 For surfaces, the total beta/gamma or alpha activity is measured by average surface contact readings using a Ludlum 43-93 radiation detector corrected for the weighted detection efficiency. To provide a conservative estimate, several assumptions are made regarding the entries in Table 2-14:

1. Surface contamination penetration is generally limited to the top 0.1 em [6]. Scan and static measurements will primarily reflect surface (or near surface) activity levels due to the limiting transmission/penetration for betas and alphas. This surface effect is reflected in the source geometry factor included in the instrument efficiency determination. (See Section 2.3.2.4, above.) The measurements of surface contamination in dpm/1 00 cm 2 have been converted to an estimated total volume activity level, considering the assumed depth (0.1 em) and instrument efficiency for the ROCs.

2. Bioshield concrete radioactivity can be attributable to neutron activation. Neutron activation levels of BWR bioshield concrete are given in Table 5.5 of NUREG/CR-3474 [14] for several depths up to 60 em in the bioshield for 30 EFPY. These levels were adjusted to reflect the VBWR operating history of 1.1 EFPY [7]. The resulting estimate of activation activity for the VBWR bioshield is 3.55E-11 Cilg, accounting for decay since shutdown. For an assumed depth of 60 em for activation products and the resulting bioshield mass, the calculated activation activity is 1.8 mCi, with an alpha component of 0.0014 mCi (based on the ratio of alpha to beta/gamma component in NUREG/CR-3474).

3. The Work Packages identified several auxiliary structures, such as grating and stairs, inside the containment building. In general, the contamination levels of such surfaces are considered lower than the surrounding building surfaces. The total area associated with such structures is a small fraction of the total surface area associated with the buildings. As a bounding estimate, the radioactivity associated with the main and basement floor areas have been increased by a factor of 1.1 to address the potential additional radioactive materials for these auxiliary structures.

Each of the above-assumed factors was chosen to be reasonably conservative based on knowledge of the facilities. The identification of the survey units and the correlation of average contamination levels to radioactivity inventory are presented in Table 2-14. The total activity associated with the VBWR being transferred to the EVESR DR-10 License is estimated to be 10.8 millicuries.

8 HITACHI TABLE 2-14:

2-23 VBWR RADIOACTIVITY INVENTORY TO BE TRANSFERRED Average Total Beta/Gamma Average Total Transuranics Area/Item Area Beta/Gamma Activity  % Alpha ~  %

Description (10 6 cm 2 )

( dpm/1 00cm 2) (pCi) Contribution (dpm/100cm 2) ~ contribution Level 1 -

Floor, Walls & I 15.6 I 21697 I 1.52E+09 I 14.10 I 7 I 4.92E+05 I 19.89 Dome Basement-Floor, Walls, 1 2.65 I 599216 I 7.15E+09 I 66.14 I 16 1.91 E+05 I 7.72

& Ceiling (less fuel Basement I 0.01 I 3346476 I 1.51 E+08 I 1.39 I 16 7.21E+02 I 0.03 Sum~

Personnel I 0.45 I 1239 I 2.53E+06 I 0.02 I 157 3.20E+05 I 12.95 Portal Equipment 0.45 4654 9.50E+06 0.09 9 1.84E+04 0.74 Portal Valve Pit 0.228 870 8.94E+05 0.01 0 O.OOE+OO 0.00 Fuel Pool 1.00 21697 9.77E+07 0.90 7 3.15E+04 1.27 BioShield 0.356 N/A 1.82E+09 16.83 N/A 1.40E+06 56.66 Embedded ,

0.100 11284 5.08E+06 0.05 3.5 1.58E+03 0.06 0.500 21697 4.89E+07 0.45 7 1.58E+04 0.64 Stack Pad 0.019 14457 1.24E+06 0.01 10 8.56E+02 0.03 Total 21.4 1.08E+10 100.0 2.47E+06 100.0 N/A- Not applicable

e HITACHI 3-1 3.0 REMAINING D&D ACTIVITIES 3.1 Introduction In accordance with 10 CFR 50.82 (a)(9)(ii)(B), the LTP must identify the major remaining dismantlement and decontamination activities. This section describes the remaining dismantlement and decontamination activities, following the guidance of NUREG-1700, "Standard Review Plan for Evaluating Nuclear Power Reactor License Termination Plans," [15]

and Regulatory Guide 1.179, "Standard Format and Content of License Termination Plans for Nuclear Power Reactors," [1]. The major dismantling activities that have been completed as of July 2023, are identified, showing the overall progress and schedule for the remaining dismantlement activities. The remaining dismantlement activities are those that remain as of July 2023.

Information is presented to demonstrate that these activities will be performed in a manner supporting a finding in accordance with 10 CFR 50.82(a)(10) that decommissioning and license termination will not be detrimental to the common defense and security or to the health and safety of the public pursuant. Information that demonstrates that these activities will not have a significant effect on the quality of the environment is provided in LTP Section 8, Environmental Report.

This section, supplemented by Section 4, identifies the major dismantlement activities that remains to be completed prior to qualifying for license termination. The final state of the VBWR will be subject to future final release under the EVESR license as part of the overall reactor facility area (300 Area). The information includes those areas and equipment in need of further remediation, and an estimate of radiological conditions that may be encountered. Included are estimates of associated occupational radiation dose and projected volumes of radioactive waste. VNC's primary goals are to decommission VBWR safely and successfully terminate the VBWR DPR-1 license. Currently, VNC has access to the radioactive waste disposal facilities in Utah and Texas for the disposal of the radioactive waste for this license termination.

VNC is currently conducting moderate decontamination and dismantlement (D&D) activities at the VBWR site in accordance with VBWR procedures and approved work packages.

Decommissioning activities at VBWR are congruent with the VNC VBWR/EVESR LPSDAR, Change Authorization Program, Radiation Protection Program, QA Program, written work plans, existing 10 CFR Part 50 license, and the requirements of 10 CFR 50.82(a)(6). If an activity requires prior NRC approval under 10 CFR 50.59(c)(2) or a change to the VBWR license conditions, a submittal will be made to the NRC for review and approval prior to implementation of the activity in question.

The discussion below provides an overview of the major remaining decommissioning activities, including activities up to future release of the site. This information includes an estimate of the quantity of radioactive material to be disposed of as radioactive waste in accordance with 10 CFR 20.2001. A description of proposed mechanisms to ensure residual radioactive materials are controlled, mitigating the potential for migration, along with estimates of occupational exposures, are included in Section 4. Characterization of radiological conditions is presented in Section 2.0. This information supports the assessment of impacts considered in other sections of the LTP and provides sufficient detail to identify inspections or technical resources needed during the remaining dismantlement activities.

Following decommissioning, radiation doses to radiation workers (for follow-up routine inspections and maintenance) will be small fractions of the limits of 10 CFR 20. For member of

e HITACHI the public, the decommissioned VBWR will not pose a potential dose exceeding the 25 3-2 mrem/year decommissioning dose criterion of 10 CFR 20.1402. Residual radioactive material will be controlled in keeping with 10 CFR 20.1406, Minimization of Contamination.

3.2 Completed Decommissioning Activities & Tasks 3.2.1 Spent Nuclear Fuel No spent nuclear fuel from the VBWR operation/license remains onsite at VNC. All VBWR fuel was transferred off site prior to 1965 per the Deactivation Report [2].

3.2.2 Liability Reduction Activities in 2007 This activity was performed under VNC Change Authorization 07-30, dated September 6, 2007.

The scope of this activity was to dismantle and dispose of remaining major components under the VBWR license with the exception of the polar crane, reactor vessel with internals, the bioshield wall (removeable), fuel pool, and the external ventilation system components.

Instrumentation and coolant components and pipes were cut and capped at the bioshield wall, packaged, and transported for radioactive waste disposal.

3.2.3 Hazard Abatement (Asbestos and Lead) and the Removeable Bioshield in 2022 The VBWR removeable bioshield, comprising mostly high-density concrete blocks with some lead bricks, was contaminated with asbestos fibers, as was the VBWR floor. A wall of asbestos insulation block was present behind the removeable bioshield and affixed to the wall. The lower-level floor of the VBWR containment was contaminated with asbestos fibers and lead dust. The asbestos insulation blocks and the asbestos and lead floor contamination were abated; associated areas were surveyed and released; the resulting hazardous waste was packaged and disposed of at a licensed disposal facility. As part of this activity, all of the removeable bioshield blocks were removed and disposed of at a hazardous waste licensed disposal facility.

The work was performed under Change Authorization CA-2022-04 dated 9/19/2022.

3.2.4 Reactor Vessel Water Removal 2022 Following reactor shutdown in 1963, the VBWR reactor vessel was stored with water inside to a level just below the recirculation nozzles, which offered radiation shielding reducing ambient area radiation levels resulting from the residual radioactive materials associated with the vessel and internals. A characterization of the vessel [7] provided information for updating radiation levels where it was determined the radiation levels to be sufficiently low precluding a High Radiation Area (HRA). In November 2022, the water was removed from the vessel to facilitate future removal.

This work was performed under Change Authorization CA-2022-02, dated 01/05/2022.

3.2.5 Reactor Vessel Pipe Removal & Closure of Openings 2023 In preparation for removing the intact reactor vessel through the 10' diameter hole in the concrete floor, all remaining reactor piping was removed to within the outer diameter of the reactor vessel flanges. This included the 10" inlet and outlet pipes, the control rod drive (CRD) nozzles and associated supports, tubing and shielding on top of the vessel, and the smaller bore thermocouple nozzles, water cooler nozzle, emergency exchanger return nozzle, and lower control nozzle. Subsequently, all remaining openings to the vessel were sealed by either

e HITACHI 3-3 welding plates over or installing blank flanges on the openings, with the exception of two of the refueling ports, which are left open to accommodate internal inspection and installation of grout, if necessary. The waste generated from this evolution will be combined with other VNC site waste and sent for disposal at a licensed disposal facility.

This work was performed under Change Authorization CA 2023-01, dated 4/24/2023.

3.3 Future Decommissioning Activities The activities remaining to be completed to support license termination are:

• Removal of the reactor vessel and its internals with its packaging and transport for permanent disposal at a commercial site.

• Decontamination of remaining radioactive materials to residual levels in keeping with the principle of ALARA; no HRAs will remain. Key areas that will be evaluated and decontaminated, as needed, include the bioshield and external sump.

• Packaging decommissioning generated material for transport and disposal as radioactive waste. It is intended that all radioactive waste from the dismantling and decommissioning activities will be transported offsite for disposal at a licensed disposal facility.

• Complete radiation surveys for remediated areas (see Section 5) and update estimates of residual radioactivity remaining (see Section 2) as needed for supporting the transfer of residual radioactivity to the EVESR license and support the requested exemption to 50.82(a)(11)(ii) in accordance with 10 CFR 50.12.

The following table lists the remaining major activities, as identified above, associated with the decommissioning of VBWR and their projected completion date:

TABLE 3-1: MAJOR REMAINING ACTIVITIES AND COMPLETION DATES Activity Projected Completion Date Reactor vessel removal Planned for 2023, NLT 2024 Ship remaining wastes to disposal 2023-2025 Perform, as needed, post-vessel removal radiation 2023-2024 surveys for previously inaccessible areas to quantify residual activity levels for transfer to EVESR Terminate the VBWR License NLT 2025 Two similar methods for removing the VBWR reactor vessel have been considered. The methodology chosen necessitates cutting an opening in the containment building and lifting the reactor vessel vertically out of the biological shield and the containment building with an external crane. The VBWR, therefore, will be removed whole through a prepared and engineered opening in the VBWR containment, will be placed on a conveyance, and transported for disposal at a licensed disposal facility and the temporary opening will be subsequently secured.

e 3.4 HITACHI Estimate of Quantity of Radioactive Material to be Shipped for Disposal 3-4 3.4.1 Solid Waste Activity & Volume Estimated volumes of waste to be generated in support of the termination of the VBWR license is provided in Table 3-2, below. The waste consists broadly of VBWR Reactor Vessel with internals and miscellaneous waste generated during dismantling activities, specifically concrete from the bioshield and piping associated with the reactor vessel.

e HITACHI 3-5 TABLE 3-2: ESTIMATED SOLID WASTE ACTIVITY & VOLUME Work Activity Volume (m 3) /(ft 3 ) Total Curies (Ci)

Rx Vessel Preparation & 97.0 /3420 36.4 Removal Ventilation system and 0 0 components Miscellaneous radioactive waste 51.0/1800 <1 (e.g., concrete, piping and DAW)

Totals 114.0/4020 37.4 3.4.2 Liquid Radioactive Waste No liquid waste is anticipated to be generated during the performance of the remaining activities other than minor quantities as may be needed to support dismantling activities. These waste waters will either be stabilized (e.g., solidified) and disposed of as radioactive waste or processed on-site at the WEP with remaining residue stabilized and disposed of as radioactive waste.

3.4.3 Airborne Radioactive Waste Activities will be performed in a manner that reduces/eliminates generation of airborne radioactive materials. In keeping with ALARA, the following processes/procedures will be used as determined appropriate in accordance with the VNC Radiation Protection program:

• Enclosures will be used during major cutting and demolition activities to prevent spread of airborne radioactive materials.

• Portable ventilation systems with HEPA filters will be used for minimizing work area airborne and limiting releases to the environment.

• Fixatives and/or protective covering will be used for preventing generation of airborne activity during movement of highly contaminated components. Dry surfaces will be wetted to minimize resuspension during demolition activities.

There are no radioactive gases at VBWR that require evaluation and control; all radioactive noble gases have decayed or previously released under appropriate control and in compliance with regulatory requirements. [2]

e HITACHI 4-1 4.0 SITE REMEDIATION PLAN 4.1 Decommissioning Objectives The primary objective is to remove the reactor vessel, which contains the majority of the remaining radioactive material; and remediate the bioshield, as necessary for meeting ALARA.

Based on current known radiological conditions of the external ventilation system, spent fuel pool and the external sump, no additional remediation is thought needed from an ALARA standpoint. Soil remediation is not anticipated as there is no soil associated with the VBWR license. Radiological remediation will be sufficient as needed for ensuring the potential radiation exposures to both workers and public from remaining radioactive materials will be ALARA with potential doses within 25 mrem/year. Residual radioactive materials will be controlled to prevent spread and contamination of the environment. (See Section 4.2, below.)

4.2 Decommissioning Activities The primary decommissioning activity is the removal of the reactor vessel. To complete this activity, supporting dismantling activities are required. These support activities include preparatory work, such as:

• Cutting an entrance in the bioshield to provide access to the reactor bottom guide (support cup) under the vessel and severing it.

• Installing bolts on two remaining open refuel ports.

• Applying flange sealant to bolted covers.

Following the removal of the reactor vessel, structures will be surveyed as necessary; contaminated materials will be remediated or removed and disposed as radioactive waste.

Residual contaminated structural surfaces that will be transferred to the EVESR license have been assessed (see Section 2). Any additional surveys necessary to determine remaining activity will be performed post dismantling I remediation. Decontamination will be performed only if determined to be ALARA for worker protection. A portion of the fixed bioshield will be removed to gain access to the reactor vessel support cup. This action may require use of diamond wire sawing and rigging activities to remove the blocks.

Appropriate decontamination techniques, equipment and materials will be identified and used, as needed, to support these and other decontamination activities. Training for workers on specialized equipment will be performed prior to use and as part of the overall work planning and control process. All work will be performed under the VNC Work Control Process, which incorporates the VNC QA Program, Radiation Protection Program and Health and Safety Program. Selection of decommissioning methods is heavily influenced by worker and public ALARA considerations. A list of VBWR facilities, planned decommissioning and decontamination activities and estimated worker exposure (person-rem) is presented in Table 4-1, below.

The following are examples of remediation techniques that may be used:

• Use of an electrically or pneumatically operated pistons with tungsten-carbide tips to fracture (scabble) concrete surface, removing a near surface layer. Exhaust air passes through both roughing and absolute high efficiency particulate air (HEPA)

e HITACHI filtration devices. Dust and debris generated through these remediation processes is 4-2 collected and controlled during the planned operations.

• Use of a needle gun, which is an electrical or pneumatic operated tool containing a series of tungsten-carbide or hardened steel rods enclosed in a housing. The rods are connected to an air-driven piston to abrade and fracture the media surface. The media removal depth is a function of the residence time of the rods over the surface.

Typically, one to two millimeters are removed per pass. Generated debris collection, transport, and dust control are accomplished in the same manner as other scabbling methods. Use of needle guns for removing and chipping media is usually reserved for areas not accessible to normal scabbling operations.

• Sponge and abrasive blasting are similar techniques that use media or materials coated with abrasive compounds such as silica sands, garnet, aluminum oxide, and walnut hulls. Sponge blasting is less aggressive, incorporating a foam media that, upon impact and compression, absorbs contaminants. The medium is collected by vacuum and the contaminants are washed from the medium so the medium may be reused. Abrasive blasting is more aggressive than sponge blasting but less aggressive than scabbling. Both operations use intermediate air pressures. Sponge and abrasive blasting are intended for the removal of surface films and paints.

• Mechanical and/or thermal techniques may be used to cutting metal structures, especially to facilitate openings in the containment to enable reactor vessel removal.

Concrete saws may be used for accurate cutting of concrete for general demolition and dismantlement. Also, this method may be used to cut large slabs of concrete for waste volume reduction or packaging. Diamond wire cutting techniques can be used to remove large segments of concrete. A diamond studded cable is circulated by a hydraulic pulley drive system through the concrete, cutting through concrete, steel rebar and other steel members in the concrete. Hydraulic cylinders control the tension of the cable. Holes are drilled through the concrete to enable stringing the cable into cutting target areas that would otherwise be inaccessible. Water applied to cool and lubricate the cable also aids in control of airborne dust. A slurry collection system is installed to collect contaminated cutting slurry, decant the slurry and recycle the water.

Remediation techniques that may be used for the structural surfaces include general washing, wiping, pressure washing, vacuuming, scabbling, chipping, and sponge or abrasive blasting.

Washing, wiping, abrasive blasting, vacuuming, and pressure washing techniques may be used for both metal and concrete surfaces. Scabbling and chipping are mechanical surface removal methods intended for concrete surfaces. Concrete removal, if required, may include using machines with hydraulic-assisted, remote-operated, articulating tools. These machines have the ability to exchange scabbling, shear, chisel, and other tool heads.

Work plans will be prepared to address the work to be accomplished and the management of the hazards involved in the relevant work scopes. The final decommissioning methods will utilize the best, most economical means to minimize hazardous, mixed and radioactive waste volume requiring licensed disposal. From the standpoint of cost-effectiveness, contaminated equipment, materials, etc., may be decontaminated, allowing release for reuse on site, or packaged for transport and disposal. This Plan allows flexibility in the choice of decontamination procedure I technique and sequence.

e 4.3 HITACHI Removal of Hazardous Materials 4-3 4.3.1 Lead All elemental lead in the form of brick has been removed and disposed of off-site. No elemental lead is expected in the areas currently inaccessible given the VBWR vessel location in the fixed bioshield. Should lead be found, it will be surveyed and/or sampled for radioactive contamination for characterization and disposal. The material will be packaged in transport containers, as necessary, removed from the containment building and eventually transported for disposal. Options for the beneficial re-use of lead will be evaluated and the most cost-effective method for final disposition pursued.

4.3.2 Lead-Containing Coatings Demolition work performed during the decommissioning activities for VBWR may require the removal of lead-containing coatings. The most likely need for abatement will be cutting of dome structures to support VBWR vessel removal. Upon identification of these areas, a qualified lead abatement subcontractor, or qualified remediation team workers, will be used as necessary to remove the lead containing coatings. Any work performed, that has a lead-containing coating, will be conducted safely by trained and qualified personnel in accordance with any applicable regulatory requirements.

4.3.3 Asbestos Abatement Asbestos containing materials have been abated in the VBWR containment and all generated material has been disposed of. Should asbestos be identified during the remaining work activities, it will be removed and packaged for disposal prior to any decommissioning activities in areas where these materials exist. Removal and disposal of asbestos will be accomplished by a licensed asbestos abatement contractor. Additional asbestos materials discovered in the course of decontamination activities will be abated by the asbestos contractor as needed.

4.4 Remediation Actions Remediation actions are performed throughout the decommissioning process. The remediation action taken is dependent on the material contaminated and extent of contamination. The principal materials that may be subjected to remediation are hardened structural surfaces and residual DAW and sludges from an ALARA standpoint. Very little decontamination, if any, is expected because contamination levels on building surfaces, the reactor vessel and related components are very low.

4.5 Decommissioning Work Controls Work controls will be established to ensure remediation work is safely performed in accordance with this LTP, VNC and VBWR license requirements, and established procedures.

Work instructions in conjunction with this LTP will be used to define the activities to be performed in the detail necessary to implement work scopes. These work instructions will be evaluated through existing VNC programs including the Change Authorization Process which incorporates 10 CFR 50.59 reviews where applicable.

Work Specifications will be developed, as needed, to establish the processes and controls needed, including use of specialized tools, methods, reviews, approvals, compliance, RP safety.

Based on work specifications, Work Packages will be prepared, containing the detailed

e HITACHI instructions for accomplishing the defined tasks. Written procedures and guidelines will be 4-4 prepared, reviewed, revised, approved, and implemented to ensure that operations are performed in a safe manner.

4.5.1 Radiation Protection Program All radiological decommissioning work activities will be performed under the VNC Radiation Protection (RP) Program, using current VNC work processes, Radiation Work Permits (RWPs),

and As Low as Reasonably Achievable (ALARA) reviews and approvals. The VNC RP Program supports all radiological operations at VNC; this program fully implements the requirements of 10 CFR 20 and has undergone years of NRC inspection. Throughout the various VNC operations, contaminated structures, systems, and components were/are decontaminated in order to perform maintenance or repair actions. The techniques used during such operations are the same or similar to the techniques used during decommissioning to reduce personnel exposure to radiation and contamination and to prevent the spread of contamination from established contaminated areas.

The performance of the dismantling and remediation activities will be evaluated, planned and controlled under the VNC Work Control Process. This process integrates a detailed examination of the hazards, risks and safety measures necessary for worker protection. The RP RWP process is an integral element of this work planning and control.

ALARA evaluations are an integral part of the RWP process. ALARA goals will be established for major dismantling activities, such as the reactor vessel removal; pre-job brief will be conducted; doses will be tracked using electronic personal (radiation) dosimeters (EPDs),

appropriate radiation safety measures, such as personal protective equipment (PPE), including respiratory protection, will be evaluated. RP job coverage is provided where needed for monitoring RP practices, controlling exposures, and minimizing dose.

The only major dismantling/decommissioning activity is the reactor vessel removal. A preliminary dose estimate has been performed for the removal and packaging of the reactor vessel. This estimate is summarized in the following table. Dose rates from gamma exposure have been estimated based on characterization surveys supplemented by the reactor vessel dose profile estimated by OW James Consulting [7]. There are no other major work activities that are anticipated to represent potentially significant radiological work. However, all work processes will be conducted under the RP Program, including pre-job dose estimates and establishing RP controls.

TABLE 4-1: A LARA DOSE ESTIMATE FOR REACTOR VESSEL REMOVAL Project Step Gamma Dose Rate Exposure Time Number Cumulative (estimate) (hours) Workers Dose Estimate (person-rem)

Install equipment 1 mrem/h (general area on 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />s/day for 20 6 0.48 (539')

for VBWR vessel 539' level); 10 mrem/h days = 160 hours0.00185 days <br />0.0444 hours <br />2.645503e-4 weeks <br />6.088e-5 months <br /> 4.8 (lower-level) removal (lower-level general area)

Lift vessel & move 1 mrem/h (general area) on 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for 1 day GA 2 0.016 (539')

to outside 539' level - middle of =8 hours, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for 0.36 Containment building, 15 mrem/hr@ 2m 3 Days @ 2m from estimated on lower VBWR vessel mid core = 12 Vessel - mid core region hours

e HITACHI 4-5 Project Step Gamma Dose Rate Exposure Time Number Cumulative (estimate) (hours) Workers Dose Estimate (person-rem)

Place vessel on 15 mrem/h @ 2m estimated 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> 2 0.09 transport on lower VBWR Vessel -

equipment mid core region Total 5.75 person-rem 4.6 Decommissioning Organization & Responsibilities The decommissioning activities will be performed under the direction and oversight of GEH and VNC site management and programs, supplemented by contracted services as needed, e.g.,

reactor vessel removal. VNC will be supported by the GE Decommissioning Organization, which has extensive experience and capability in remediation techniques on contaminated systems, structures, and components during decommissioning. Staff will be trained as required to meet site and regulatory requirements. GEH VNC will conduct routine meetings for work planning, including radiation protection and ALARA, and Safety reviews of decommissioning activities to ensure site goals and expectations are met.

4.6.1 Quality Assurance Quality Assurance Decommissioning will be performed under the existing NRC approved GE Hitachi Nuclear Energy Americas Quality Assurance Program Description (NED0-11209-A, Revision 17) [16] (Note- NRC approved this document on 1/22/21. NED0-11209 replaced the VNC Quality Assurance Plan (QAP)). A Decommissioning Quality Assurance Project Plan will be developed, as needed, for defining and/or expanding the quality requirements for the decommissioning efforts.

4.6.2 Contractor Assistance GEH management may select qualified contractors to assist in selected, specialized activities, such as the VBWR Reactor Vessel removal and disposal at a licensed disposal facility. Other elements of the project may entail the use of subcontracted employees for activities such as radiological controls, radiological surveys, structural analysis, concrete cutting, etc. All activities will be evaluated and controlled under the VNC programs, as discussed above. GEH will be in charge of the overall project management, engineering, radiological controls/surveys, waste management and shipping and general oversight of all subcontractors. Other contractors may be added to the team as needed throughout the project. Contractors and subcontractors performing work under this LTP will be required to comply with the applicable GEH site policies and procedures.

e HITACHI 5-1 5.0 FINAL STATUS SURVEY A request pursuant to 10 CFR 50.82(a)(11), for an exemption from 10 CFR 50.82(a)(11) requiring that the Vallecitos Boiling Water Reactor (VBWR) meet the requirements for release according to 10 CFR Part 20, subpart E, has been submitted by GE-Hitachi Nuclear Energy Americas LLC (GEH) (Ref the transmittal letter). In support of that exemption request, this transmittal also includes a license amendment request to transfer the remaining VBWR facility and its in-situ residual radioactivity to the authority of the ESADA Vallecitos Experimental Superheat Reactor (EVESR) DR-10 license. These actions are according to the decommissioning licensing strategy first outlined in GEH Letter to the NRC, GEH Description of Process for Decommissioning the Vallecitos Nuclear Center Shutdown Reactors (ADAMS Accession No. ML21315A005), November 11, 2021, and later developed in the VNC Limited Post Shutdown Decommissioning Activities Report (LPSDAR) (ADAMS Accession No. ML22264A325), September 21, 2022. This series of actions will enable VBWR DPR-1 to be terminated so that its final decommissioning can be accomplished coincident with the final decommissioning of the EVESR facility and the three decommissioning reactors at the Vallecitos Nuclear Center (VNC); VBWR, EVESR, and the GE Test Reactor (GETR) may undergo a holistic final decommissioning based on the EVESR final license termination date of April 15, 2030.

e HITACHI 6-1 6.0 COMPLIANCE WITH THE RADIOLOGICAL CRITERIA FOR LICENSE TERMINATION A request, pursuant to 10 CFR 50.82(a)(11), for an exemption from 10 CFR 50.82(a)(11),

requiring that the Vallecitos Boiling Water Reactor (VBWR) meet the requirements for release according to 10 CFR Part 20, subpart E, has been submitted by GE-Hitachi Nuclear Energy Americas LLC (GEH) (Ref the transmittal letter).

A final status survey plan is to be addressed as part of the decommissioning activities for EVESR. This plan, addressing the license termination for EVESR and release of the 300 Area will meet the requirements of 10 CFR 50.82(a)(11) for EVESR and the remaining, transferred VBWR facilities. Accordingly, this plan will demonstrate that the dose from residual radioactivity that is distinguishable from background radiation does not exceed 25 millirem (mrem) (0.25 millisievert (mSv)) per year to an average member of the critical group from all appropriate pathways over a 1,000-year period. This final status survey plan will be submitted in its entirety as a component of the EVESR License Termination Plan prior to April 15, 2028.

e HITACHI 7-1 7.0 FINANCIAL GEH has established a decommissioning Financial Assurance Mechanism in the form of payment surety bonds for each of the GEH facilities that hold NRC licenses. The Financial Assurance Mechanism established by this approach meets all the requirements of the NRC's decommissioning financial assurance regulations contained in 10 CFR 50. Appropriate updates have been submitted to the NRC to maintain adequate levels of financial assurance.

In March of 2023, GEH submitted to the NRC surety bond riders updating the decommissioning cost estimates for each of the VNC NRC licenses (ADAMS Accession No. ML23104A417) [17]

Future updates will be made as appropriate.

e HITACHI 8-1 8.0 ENVIRONMENTAL REPORT GEH has prepared an Environmental Report (ER) [18] to evaluate the potential radiological and non-radiological impacts associated with the removal and disposal of the VBWR vessel in the 300 Area at VNC. The attached ER (Attachment 1) is submitted in accordance with the requirements of 10 CFR 50.82(a)(9) and 10 CFR 51.53(d) to address the post-operating license stage of the facility. As required by these regulations this ER addresses new information and environmental changes associated with the proposed termination activities.

The evaluation led to the following conclusions:

1. The technology for decommissioning nuclear facilities is established and while technical improvements in decommissioning techniques are to be expected, decommissioning at the present time can be performed safely and at a reasonable cost.

2. Decommissioning nuclear facilities is not an imminent health and safety problem.

3. Decommissioning of a nuclear facility generally has a positive environmental impact. Since in many instances, such as at a reactor facility, the land is a valuable resource, return of this land to the commercial or public sector is highly desirable.

In summary, the evaluation of potential environmental effects upon each of the respective resource areas by the proposed action is small. In accordance with the NRC's established standard, the environmental effects are not detectable or are so minor that they would neither destabilize nor noticeably alter any important attribute of the resource.

e HITACHI 9-1 9.0 BIBLIOGRAPHY

[1] NRC, "Regulatory Guide 1.179, Standard Format and Content of License Termination Plans for Nuclear Power Reactors, Rev. 2," 2019.

[2] G. E. C. "Final Report on Deactivation of Vallecitos Boiling Water Reactor, Docket 50-18,"

February 5, 1965.

[3] ReNuke/ENERCON, "REN-TR-22-011, SDR Radiological Characterization Report, Rev.

1,"March 2023.

[4] ReNuke/ENERCON, "Vallecitos Nuclear Center Decommissioning Radiological Seeping Summary Report".

[5] ReNuke/ENERCON, "REN-TR-22-010, Vallecitos Nuclear Center Decommissioning Historical Site Assessment, Rev. 0," October 2022.

[6] NRC, "NUREG-1575, Rev. 1, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM)," 2000.

[7] D. W. James Consulting, "Vallecitos Nuclear Center VBWR Reactor Vessel and Internals Characterization and Classification - 2022," March 2022.

[8] Chesapeake Nuclear Services, Inc., "Radiological Assessment of Outside Contaminated Areas at the Vallecitos Nuclear Center," 2020.

[9] B. &. Caldwell, "Letter Report from Brown and Caldwell to Zanotto, GE-Hitachi, dated July 22, 2019," 2019.

[10] ReNuke/ENERCON, "REN-TR-003, Vallecitos Nuclear Center Decommissioning Industrial Hygiene Seeping Report, Rev. 2," October 2022.

[11] ReNuke/ENERCON, "REN-TR-22-008, Closeout Report for Asbestos and Lead Abatement, Rev. 0," December 2022.

[12] ReNuke/ENERCON, "CP-VNC-RADChar-00007, Vallecitos Nuclear Center Decommissioning Radiological Characterization Plan, Rev. 0," April 2022.

e HITACHI 9-2

[13] NRC, "NUREG-1507, Rev. 1, Minimum Detectable Concentrations with Typical Radiation Survey for Instruments for Various Contaminants and Field Conditions," U. S. Nuclear Regulatory Commission, August 2020.

[14] NRC, "NUREG/CR-3474, Long-lived Activation Products in Reactor Materials," NRC, 1984.

[15] NRC, "NUREG 1700, Standard Review Plan for Evaluating Nuclear Power Reactor License Termination Plans, Rev 2," 2018.

[16] GE-Hitachi, "NED0-11209-A, GE Hitachi Nuclear Energy Quality Assurance Program Description, Rev. 16," 2022.

[17] S. P. Murray, "Letter M230040, GNF-A/GEH Financial Assurance of Decommissioing Funds - Surety Bond Riders," March 31, 2023.

[18] I. (. Surf to Snow Environmental Management, "Vallecitos Boiling Water Reactor, Environmental Report; Initial Report- 08-18-23," San Ramon, 2023.

[19] Stantec Consulting Services, Inc., Phase I Environmental Site Assessment for the Vallecitos Site, Sunol, California, 2019.

8 HITACHI 9-1 Attachment 1 - VBWR Environmental Report