Evaluation of Radiological Protection in Hot Cell Facility during Processing Decommissioning Radioactive Waste

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Master's Thesis

Evaluation of Radiological Protection in Hot Cell Facility during Processing Decommissioning Radioactive Waste

from Pressurized Water Reactor

Kyung Hun Shin

Department of Nuclear Engineering

Ulsan National Institute of Science and Technology

2023

Evaluation of Radiological Protection in Hot Cell Facility during Processing Decommissioning Radioactive Waste

from Pressurized Water Reactor

Kyung Hun Shin

Department of Nuclear Engineering

Ulsan National Institute of Science and Technology

Evaluation of Radiological Protection in Hot Cell Facility during Processing Decommissioning Radioactive Waste

from Pressurized Water Reactor

A thesis/dissertation submitted to

Ulsan National Institute of Science and Technology in partial fulfillment of the

requirements for the degree of Master of Science

Kyung Hun Shin

Kyung Hun Shin

This certifies that the thesis/dissertation of Kyung Hun Shin is approved.

12.06.2022 of submission

Evaluation of Radiological Protection in Hot Cell Facility during Processing Decommissioning Radioactive Waste

from Pressurized Water Reactor

ABSTRACTS

After the Fukushima nuclear accident in 2011, radioactive waste management measures and radioactive waste from nuclear power plants are attracting attention worldwide. In Korea, radioactive waste characterization and management measures are required as Kori Unit 1 is permanently shut down in 2017 and proceeds with an immediate decommissioning strategy. When decommissioning a nuclear facility, contaminated radioactive metals and concrete will account for most of the decommissioning radioactive waste. An unknown amount of radioactive waste including nuclides produced by unidentified impurities was produced during the decommissioning procedure required for the Kori Unit 1 nuclear power plant's (NPP) permanent shutdown. Therefore, when decommissioning a nuclear power plant, it is essential to predict the exposure dose of workers for radiation protection and to secure safety according to the radioactivity concentration and air pollution monitoring plan in the working area. To ensure such safety, analysis of the characteristics of decommissioning radioactive waste should be performed in a hot cell facility. Therefore, in this study, safety evaluation such as safety evaluation of hot cell facilities, workers' exposure dose evaluation, and radioactive material emission characteristics were performed in this process of handling highly uncertain waste. Additionally, the behavior of the airborne source term that occurs during the processing of radioactive waste when ventilation is considered in the hot cell facility was examined. The radioactive waste to be dealt with in the hot cell facility is composed of carbon steel and stainless steel as the reactor pressure vessel and reactor internal structure extracted from the reactor. More than 99% of the radioactivity inventory of the reactor pressure vessel and core barrel of the Pressurized Water Reactor (PWR) was composed of six nuclides; ⁶⁰Co,

63Ni, ⁵⁵Fe, ³H, ⁵⁹Ni, and ¹⁴C. Additionally, the Monte Carlo N-Particle (MCNP) transport code was used to evaluate the workers' external exposure dose in the event of an accidental ventilation failure, a broken window, and normal operations at the hot cell facility during the processing of radioactive waste. The case where the highest dose value was obtained was evaluated when the reactor pressure vessel was processed with a circular saw. When the core beltline of reactor pressure vessel was cut using a circular saw, the case where the highest dose value was obtained was evaluated. The external exposure dose in the event of an unexpected ventilation failure was calculated at 2,950 times that of normal operation, and the external exposure dose because of a broken lead glass window was calculated as an additional 140 times. It was confirmed that the external exposure dose from the case of broken lead glass was 4.12E+05 times higher than the exposure dose from normal operation. Under the assumption that 1%

of the period of the airborne source term will leak during normal operation, the internal exposure dose of the worker was estimated to be 1.84 × 10^-2 mSv/y, the highest in the same case. When worker exposed to contaminated air, the worker's exposure dose complied with the dose limit since it was confirmed that the worker was exposed by inhalation. In addition, the internal exposure dose evaluation was evaluated under the assumption that the airborne source term would flow into the working area in the

event of an earthquake accident. At this time, the dose rate in the case of the highest dose value on the worker is 4.37 × 10^-1 mSv/h. According to Article 14 of the Nuclear Safety and Security Commission Notification No. 2019-10, it was possible to infer the time to secure the safety of the worker. This enabled it to predict how long an irradiated worker could have access to their workspace prior to an accident. At this case, emergency working hours are allowed for a minimum of 1,145.47 hours, which should be sufficient to keep down the accident. Accordingly, a high efficiency particulate air filter (HEPA) with an efficiency of 99.97% was used to detect the emission concentration of the dominant nuclides (³H, ¹⁴C, ⁵⁵Fe, ⁵⁹Ni, ⁶⁰Co, and ⁶³Ni) of the airborne source term, which complies with the emission management standard of Nuclear Safety Commission Notice No. 2016-16. Based on these results, it is possible to evaluate the emission management in the airborne source term while processing the radioactive waste in a hot cell facility and performing radiation protection technology.

１

CONTENTS

I. Introduction ··· 6

II. Literature Study ··· 8

2.1 Decommissioning ··· 8

2.2 Decommissioning strategy ··· 9

2.3 Status of decommissioning at domestic and abroad ··· 10

2.4 Decommissioning process according to regulation ··· 12

2.5 Generation and types of decommissioning radioactive waste ··· 16

2.6 Regulation measure of radioactive waste ··· 17

2.7 Physicochemical properties of decommissioning waste ··· 19

2.8 Hot cell facility ··· 20

2.9 Post irradiation examination facility ··· 21

2.10 Radioactive waste handling process in hot cell facility ··· 22

2.11 Radiation work and dose regulation measures ··· 24

III. Materials and Methods ··· 25

3.1 Basic design of hot cell facility ··· 25

3.2 Decommissioning radioactive waste from PWR ··· 27

3.3 Nuclide inventory evaluation of decommissioning radioactive waste ··· 29

3.5 Aerosol production during cutting process ··· 33

3.6 Aerosol mass ··· 34

3.7 Main routes of exposure for radiation workers ··· 35

3.8 Airborne source term concentration ··· 36

3.9 Indoor particle concentration ··· 37

3.10 MCNP evaluation methodology··· 38

IV. Results and discussions ··· 39

4.1 Environment radiation emission evaluation ··· 39

２

4.2 Dose evaluation of MCNP simulation ··· 50

4.3 Radioactive concentration evaluation of working area ··· 52

4.4 Internal dose evaluation of normal operation ··· 59

4.5 Internal dose evaluation of accident scenarios ··· 63

V. Conclusion ··· 67

REFERENCES ··· 68

ACKNOWLEDGEMENT ··· 70

３

LIST OF FIGURES

Figure 1. Radioactive waste classification [1] ... 6

Figure 2. Sampling and processing of decommissioning waste ... 7

Figure 3. Kori Unit 1 of NPP ... 8

Figure 4. Steps to perform decommissioning ... 13

Figure 5. Concept of regulation management for radioactive materials ... 18

Figure 6. Activation of PWR vessel cladding type 304 stainless steel [5] ... 19

Figure 7. Irradiated Materials Examination Facility ... 20

Figure 8. 3D view of PIEF 9405 ... 21

Figure 9. 3D view of AHCF ... 25

Figure 10. Typical Pressurized Reactor ... 28

Figure 11. Geometry of considering ventilation and non-ventilation in AHCF ... 38

Figure 12. Core beltline metal aerosol of Cin ... 46

Figure 13. Core weld metal aerosol of Cin ... 46

Figure 14. Core barrel metal aerosol of Cin ... 47

Figure 15. Dose rate of processing with core beltline in AHCF ... 50

４

LIST OF TABLES

Table 1. Decommissioning completion status [2] ... 11

Table 2. Confirmation of decommissioning status of nuclear facility∙subject to inspection [3] ... 15

Table 3. Estimated amount of NPP decommissioning waste [4] ... 16

Table 4. Dose Limit of Enforcement Decree in the Nuclear Safety Act [11] ... 24

Table 5.Nuclide inventory of core beltline (reactor pressure vessel) [14] ... 30

Table 6. Nuclide inventory of core barrel (reactor internal structure)[14] ... 30

Table 7.Data on decommissioning waste sample volume, amputated sample volume, and cutting loss volume ... 32

Table 8.Data of core beltline metal aerosol with circular saw ... 40

Table 9. Data of core weld material aerosol with circular saw ... 41

Table 10. Data of core weld material aerosol with circular saw ... 42

Table 11. Data of core weld material aerosol with band saw ... 43

Table 12. Data of core barrel metal aerosol with circular saw ... 44

Table 13. Data of core barrel metal aerosol with band saw ... 45

Table 14. Emission management standard [21] ... 48

Table 15. Dose rate during normal operation ... 51

Table 16. Dose rate during accidental fire ... 51

Table 17. Dose rate in an earthquake accident ... 51

Table 18. Data of core beltline aerosol with circular saw in normal operation ... 53

Table 19. Data of core beltline aerosol with band saw in normal operation ... 54

Table 20. Data of core weld metal aerosol with circular saw in normal operation ... 55

Table 21. Data of core weld metal aerosol with band saw in normal operation ... 56

Table 22. Data of core barrel metal aerosol with circular saw in normal operation ... 57

Table 23. Data of core barrel metal aerosol with circular saw in normal operation ... 58

Table 24. Inhalation effective dose coefficient ICRP publication 68 ... 59

５

Table 25. Internal dose rate during processing core beltline with circular saw in normal operation .... 60

Table 26. Internal dose rate during processing core beltline with band saw in normal operation ... 60

Table 27. Internal dose rate during processing core weld with circular saw in normal operation ... 61

Table 28. Internal dose rate during processing core weld with band saw in normal operation... 61

Table 29. Internal dose rate during processing core barrel with circular saw in normal operation ... 62

Table 30. Internal dose rate during processing core barrel with band saw in normal operation ... 62

Table 31. Internal dose value during processing core beltline with circular saw in earthquake accident ... 64

Table 32. Internal dose value during processing core beltline with band saw in earthquake accident . 64 Table 33. Internal dose value during processing core weld with circular saw in earthquake accident . 65 Table 34. Internal dose value during processing core weld with band saw in earthquake accident ... 65

Table 35. Internal dose value during processing core barrel with circular saw in earthquake accident 66 Table 36. Internal dose value during processing core barrel with band saw in earthquake accident .... 66

6

I. Introduction

After the Fukushima nuclear power plant accident in 2011, the management of radioactive waste generated from nuclear power plants and the decommissioning of nuclear facilities were emphasized.

Korea’s first pressurized light water reactor (PWR), Kori Unit 1 of nuclear power plant (NPP), started commercial operation in 1978, finished its life extension in 2017, and is preparing immediate decommissioning strategy. In addition, a total of 12 nuclear power plants in Korea are expected to reach the end of their lifespan by 2030, and evaluation of radioactive properties of radioactive waste has become an essential process. According to Article 2 of the Nuclear Safety Act, radioactive waste is defined as radioactive material or material contaminated by it and is subject to disposal. Radioactive waste is classified into high-level, low-intermediate-level, and very low-level clearance radioactive waste according to the concentration of radioactivity. High-level radioactive waste is classified as spent nuclear fuel used in nuclear reactors. Medium-low-level radioactive waste is classified as radioactive isotope (RI) waste generated by hospitals, research institutes, universities, and industries, such as devices and parts used in nuclear power plants and work clothes for radiation workers. In addition, very- low level radioactive waste is classified as radioactive waste that is less than 100 times the allowable concentration for its own disposal. Since the metal radioactive waste generated during the decommissioning process of nuclear power plants contains nuclides generated by impurities, it is necessary to understand the characteristics of the decommissioned radioactive waste and secure safety during the disposal process.

Figure 1. Radioactive waste classification [1]

7

For this process, it is necessary to extract samples from nuclear reactors, which are core components of nuclear power plants, and then process them directly at hot cell facilities to understand the characteristics of the extracted radioactive waste. Because metal radioactive waste is processed inside the hot cell, the evaluation of the radiation safety technology of the hot cell facility is an essential process. The hot cell facility handles radioactive waste and consists of a shield wall and lead glass for the safety of workers. In such a thermal cell facility, radiation safety evaluation by workers and environmental impact evaluation by filter systems are essential. Decommissioning waste treated in hot cell facilities is metal waste, which consists of core beltline extracted from the reactor pressure vessel, core weld, and core barrel in the reactor internal structure. The core beltline and the core weld are made of carbon steel, and the core barrel is made of SUS304 stainless steel. When the decommissioning waste sample collected in this way is processed with a circular saw or a band saw, the cutting loss volume by the saw is generated, some of which is generated in the form of aerosol. Since this aerosol can cause internal exposure by human respiration, the radiation safety examination of radiation workers based on ventilation and leakage rate of hot cell facilities was evaluated through MCNP code to determine whether workers' doses meet regulatory standards. Also, by evaluating the concentration of nuclides released during aerosol discharge by the ventilation system, it was evaluated whether the emission control regulatory standards under the Nuclear Safety Act were satisfied.

Figure 2. Sampling and processing of decommissioning waste

8

II. Literature Study

2.1 Decommissioning

Decommissioning refers to all technical and administrative activities to permanently remove a nuclear facility that has lost its utility value from the surrounding environment. According to Article 2, Paragraph 24 of the Nuclear Safety and Security Act, "decommission" refers to all activities to be excluded from the application of this Act by permanently suspending the operation of power generation reactors, research, educational reactors, and nuclear fuel cycle projects. According to Article 2, Paragraph 24 of the Nuclear Safety and Security Act, 'decommissioning' means permanently suspending the operation of facilities licensed for power generation reactors, educational reactors, and nuclear fuel cycle projects, then dismantling the facilities and sites or removing radioactive contamination. By doing so, it refers to all activities to exclude from the subject of this law. Because this is premised on securing the safety of residents and the surrounding environment, the International Atomic Energy Agency (IAEA) and Nuclear Energy Agency (NEA) classify the following types of decommissioning according to strategic aspects, which is called decommissioning strategy [1].

Figure 3. Kori Unit 1 of NPP

9

2.2 Decommissioning strategy

2.2.1 Immediate decommissioning

For facilities that handle radionuclides with very long half-lives, such as reprocessing facilities or nuclear fuel manufacturing plants, it is a decommissioning strategy that is applied immediately after the operation of nuclear facilities is stopped.

2.2.2 Delayed decommissioning

Delayed decommissioning is a strategy to keep a facility or site safe for a period and then decontaminate and decommission it. A monitoring and maintenance program is implemented to ensure that the required level of safety is maintained during the deferred decommissioning period.

2.2.3 Entombment

Long-term structurally and chemically stable materials such as concrete are permanently closed, and recently their feasibility is being reviewed in research reactors with very complex and small-scale structures. A typical example is the decommissioning strategy applied to the Chernobyl NPP.

10

2.3 Status of decommissioning at domestic and abroad

In 2021, a total of 443 nuclear power plants are in operation in 38 countries around the world, and a total of 193 nuclear power plants are permanently shut down. In the United States, 40 nuclear power plants were permanently shut down, 16 of which were decommissioned, completed 3 of 30 in Germany, only shunt down 30 in The United Kingdom, completed 1 of 27 in Japan, only shut down 14 in France, and 1 in Switzerland and 2 in Korea were permanently shut down. In Korea, there is no decommissioning of commercial NPP other than permanent nuclear reactors, and it is required to evaluate radioactive contamination and secure radioactive waste handling technologies, which are key elements of nuclear decommissioning technology [1].

11

Table 1. Decommissioning completion status [2]

Nation Name of NPP Reactor Type

Generation Capacity

(Mwe)

Operating Year

Suspended Year

Decommissioning Completed Year

U.S.A

BIG ROCK POINT BWR 67 1962 1997 2007

BONUS BWR 17 1964 1968 1970

CVTR PHWR 17 1963 1967 2009

ELK RIVER BWR 22 1963 1968 1974

FORT ST. VRAIN HTGR 330 1976 1989 1996

HADDAM NECK PWR 560 1967 1996 2008

HALLAM - 75 1963 1964 1971

MAINE YANKEE PWR 860 1972 1997 2005

PATHFINDER BWR 59 1966 1967 2007

Rancho Seco PWR 918 1975 1989 2009

PIQUA - 12 1963 1966 1969

SAXTON PWR 3 1967 1972 2005

SHIPPINGPORT PWR 60 1957 1982 1990

SHOREHAM BWR 820 1986 1989 1995

TROJAN PWR 1,095 1975 1992 2005

YANKEE NPS PWR 167 1960 1991 2005

Germany

GROSSWELZHEIM BWR 25 1969 1971 1998

NIEDERAICHBACH HWGCR 100 1973 1974 1995

VAK Kahl BWR 16 1961 1985 2010

Japan JPDR BWR 12 1963 1976 1996

Switzerland LUCENS HWGCR 6 1968 1969 2004

Total 21

12

2.4 Decommissioning process according to regulation

2.4.1 Preparation step for decommissioning NPP

The first step in the decommissioning process is to determine the decommissioning specifications for decommissioning a nuclear facility at the end of its useful life. Since the basic specifications for decommissioning have a much influence on the economy and safety of the decommissioning process, the process of establishing a decommissioning strategy is a very important process. In fact, since regulatory actions that deviate from the original purpose constitute regulatory failure, preemptive safety rules and the clarity of safety rules should be observed. The basic specifications for decommissioning are the scope of decommissioning, the time and method of decommissioning, the final status of the decommissioning site, evaluation and setting of decommissioning costs, and the final disposal plan for waste. Also, securing the technology for this includes radiation safety management, decommissioning waste standard setting, and decommissioning waste management policy. When the basic specifications for decommissioning are decided, the physicochemical and radiological characteristics investigations of nuclear facilities are carried out. Investigation of the radioactive properties of nuclear facilities is the most important task to be performed prior to decommissioning and is the process carried out in this study. This characteristic evaluation determines major decommissioning scenarios, such as the need for shielding or isolation facilities, based on the basic information of radiation safety analysis such as environmental impact assessment and worker exposure dose analysis [1]. The facility characteristic evaluation will investigate the distribution of external doses, pollution and contamination within the decommissioning facility, the type of radionuclides, and the amount of radioactive inventory. Based on the information investigated in this way, the decommissioning plan is submitted to the regulatory institute, which is classified as a document that can be approved by the regulatory institute that the decommissioning process can be completed safely and efficiently. For decommissioning design and accreditation of permanently stopped nuclear facilities, detailed plans for decommissioning, forecasting of supply and demand for manpower and materials, decommissioning project design, industrial safety evaluation of decommissioning, environmental impact assessment and establishment of countermeasures are made [1]. In addition, it is essential to predict and minimize workers' exposure dose to protect radiation when decommissioning, and to secure safety according to the spatial radiation dose and air pollution monitoring plan in the workplace [1].

13

Figure 4. Steps to perform decommissioning Decommissioning Planning

Suspension of operation Analysis of facility

characteristics

Decommissioning Design

Decontamination, Cutting, Demolition, Waste management

Restoration of the site

Clearance of the site

14 2.4.2 Decommissioning NPP by Nuclear Safety Act

In 2011, IAEA requested the Korean government to establish and periodically renew a plan for decommissioning nuclear facilities [2]. Accordingly, when obtaining a construction permit for a nuclear facility, a preliminary decommissioning plan must be submitted to the regulatory institute for approval.

Articles 10 (construction permit) and 20 (operation permit) of the Nuclear Safety Act suggest that changes in the accreditation document are submitted for approval at the time of operation permission.

This preliminary decommissioning plan presents the nuclear decommissioning strategy, schedule, radiation disaster prevention measures and technical standards related to decontamination by radioactive materials in Article 85 paragraph 7 of the Nuclear Safety Act. In addition, Articles 28 (decommissioning of nuclear power reactors and related facilities) and 42 (decommissioning of nuclear fuel cycle facilities) of the Nuclear Safety Act suggest that the final decommissioning plan should be approved, submitted, decommissioning completed report checked, and inspected. In accordance with Article 41 paragraph 2 of the Enforcement Decree of the Nuclear Safety Act, regulatory institute first stipulate that the technical capabilities required for decommissioning plans meet the technical standards and that the exposure dose due to decommissioning is below the dose limit. In addition, Article 28 paragraph 2 of the Nuclear Safety Act proposes to submit a quality assurance plan, including a final decommissioning plan describing the approval requirements, residents' opinions on the draft decommissioning plan, and the results of the public hearing [2]. After approval for decommissioning is granted, the nuclear operator must report on the decommissioning status. Accordingly, the Nuclear Safety and Security Commission conducts on-site verification and inspection of this, in accordance with Nuclear Safety Commission Notification No. 2016-32 on how to check and inspect the status of decommissioning of nuclear facilities [3]. Decommissioning status inspection inventory include decontamination, decommissioning, radiation, environmental radiation, radioactive waste, fire protection, and product warranty [3]. Accordingly, it presents laws and technical standards related to nuclear safety regulations by preparing facilities subject to decommissioning, updating preliminary decommissioning plan, approving decommissioning, checking, and inspecting decommissioning, completing decommissioning, and terminating permits. Since some of the radioactive waste generated after the decommissioning of nuclear power facilities can be self-disposal through regulation clearance, it is possible to effectively treat radioactive waste generated during the decommissioning process of nuclear power plants through sampling, nuclide analysis, and radioactivity measurement processes [2].

15

Table 2. Confirmation of decommissioning status of nuclear facility∙subject to inspection [3]

Check verification subject

field Check- Verification

Decontamination activation - Decontamination of Structure, system and equipment - Soil, surface and ground water restoration

Decommissioning activation - Decommissioning of structures, systems and equipment

Radiation management - Radiation safety and radiation protection activities - Radiation monitoring and measurement

Environment radiation management

- Environment radiation management - Laboratory product management

Radioactive waste management

- Liquid radioactive waste management - Gas radioactive waste management - Air ventilation system

- Solid radioactive waste management

- Temporary storage facility for radioactive waste

Fire protection - Fire protection facilities and fire protection activities Product warranty - Matters described in the quality assurance plan for

decommissioning

16

2.5 Generation and types of decommissioning radioactive waste

Decommissioning radioactive waste means radioactive materials generated during the decommissioning of nuclear facilities or materials polluted by them and subject to disposal. The generation of decommissioning waste occurs at all stages of decontamination, cutting, and environmental restoration for the decommissioning of nuclear facilities, and among the wastes generated during facility operation, the remaining untreated/disposable wastes are also regarded as decommissioning wastes at the time of decommissioning. A characteristic of decommissioning waste is that there are various types and various levels of radioactivity. because solid waste is much more than waste generated during the operation of nuclear power facilities and there are various types, materials, and forms [4]. The types of decommissioning waste include special waste such as metal waste, concrete waste, contaminated lead, graphite, sludge, and mixed organic/inorganic waste. It is also classified as combustible waste such as protective clothing and wood, and waste during operation that is not treated and stored, such as waste ion exchange resin. Metal and concrete wastes account for the majority of nuclear power plant decommissioning, and among them, metal waste consists of primary systems such as structures in large reactors and pressure vessels [4]. The amount of decommissioning waste is not easy to predict with a generalized method, and the amount of decommissioning waste is determined by the operation history, the decommissioning strategy of the nuclear power plant operator, or the national radioactive waste management regulations.

Table 3. Estimated amount of NPP decommissioning waste [4]

Type of waste

GCR (250 Mwe) PWR (900-1300 Mwe)

Weight (ton) Weight (ton) Volume (m³) Drum (number)

Radioactive metal 3,000 650 93 546

Graphite 2,500 - - -

Radioactive concrete 600 300 150 882

Contaminated metal 6,000 3,500 500 2,941

Contaminated

concrete 150 600 300 1,765

Contaminated

finishing material 150 150 21 126

Decommissioning

dry active waste 1,000 1,000 1,250 7,353

Total 13,400 6,200 2,314 13,613

17

2.6 Regulation measure of radioactive waste

In all cases where radioactive materials are used, stored, or disposed of each process must be carried out in a defined by the laws of the under nation, and all procedures must be reported or approved by the regulatory institute as stipulated. According to the Nuclear Safety Commission Notification No. 2020- 6, radioactive waste classification and self-disposal standards, radioactive wastes that have been proven to satisfy the permissible dose for self-disposal and those presented in Article 107, Paragraph 3, Item 2 of the Nuclear Safety Commission Notification Self-Disposal. It is prescribed that radioactive waste below the allowable concentration can be disposed of by itself. The basic specifications for decommissioning are the scope of decommissioning, the time and method of decommissioning, the final status of the decommissioning site, evaluation and setting of decommissioning costs, and the final disposal plan for waste.

2.6.1 Exclusion

It refers to the case where the radiation source itself is naturally occurring and there is no way to regulate it, and it is excluded from the target of regulation.

2.6.2 Exemption

This refers to the case where a specific activity is below the prescribed value and is not subject to notification or permission because it is not subject to regulation. There is no need for operators or owners to register use and ownership with regulatory institute.

2.6.3 Clearance

It means that the registered regulated material is out of the regulatory system, and for deregulation, the radiological effect of deregulation of radioactive materials or waste must be evaluated. It is much smaller than the standard for exemption from regulation.

18

Figure 5. Concept of regulation management for radioactive materials

Regulation exempt

Source/Activity/Exposure

Permission emission

Limited clearance

Permanent disposition

Regulation management probability?

Minor harm under limited condition?

Minor harm under

all conditions?

Regulation management

Minor harm under

limited condition?

Minor harm under

limited condition?

Minor harm under all conditions?

Unlimited Clearance Yes

Yes

Yes Yes

No

No

No

No No

Regulation exempt

Regulation exempt

Yes

Yes

19

2.7 Physicochemical properties of decommissioning waste

Decommissioning waste is largely divided into metal structures and concrete, and since most large structures around the core are made of stainless steel (SUS 304L), the radioactive properties of metal Decommissioned wastes were targeted as stainless steel and investigated. The main radioactive nuclides of stainless steel are ⁶⁰Co, ⁵⁵Fe, ⁵⁴Mn, and ⁶³Ni, and ⁶⁰Co is the most dominant nuclide until about 10 years after decommissioning. Although cobalt is not a basic constituent element, ⁶⁰Co is produced through the ⁵⁹Co(n,γ)⁶⁰Co reaction even if it is contained as an impurity and contains a trace amount [5].

The uncertainty of the cobalt impurity content of stainless steel is very large, and if all radioactivity is evaluated based on ⁶⁰Co, a conservative evaluation is possible [5].

Figure 6. Activation of PWR vessel cladding type 304 stainless steel [5]

20

2.8 Hot cell facility

Hot cell facility is a facility that can safety and remotely handle radioactive materials and consists of a concrete and lead glass that can shield radiation. Through hot cell facilities, nuclear fuel-related performance, safety demonstration tests, nuclear fuel development, and performance-related research tasks are supported. Because the hot cell facility is divided into radiation-controlled area, when radiation workers, frequent visitors, and temporary visitors entering and exiting this facility are managed separately. Temporary visitors must wear thermos luminescence dosimeters (TLD) or auxiliary dosimeters. In addition, for the contamination management in the radiation-controlled area, the design dose rate of the hot cell facility must be managed within the dose limit. According to article 3 of the rules on the technical standards for radiation protection, etc., the regulatory institute is reviewing whether the external radiation dose rate is maintained within 400 μSv per week.

Figure 7. Irradiated Materials Examination Facility

21

2.9 Post irradiation examination facility

Post irradiation examination facility (PIEF) is the hot cell facility in Korea atomic energy research institute (KAERI) that can handle the entire field of spent nuclear fuel (SNF) assembly for PWR. This Facility is contributing to securing the soundness of domestic nuclear power plant operation and developing nuclear fuel localization technology. PIEF consists of three water tank facilities that can handle the entire length of the spent fuel assembly for PWR, and number of six hot cell facilities that can perform non-destructive and destructive tests of spent nuclear fuel rods. Among the six hot cell facilities, the PIEF 9405 of concrete hot cell facility performs fission gas capture and fuel rod cutting.

The PIEF 9405, which has an internal volume of 4.0 × 1.5 × 3.5 m³, is depicted in Figure 8, working aera is 19.8 × 5.5 ×5 m³ [6]. The maximum allowable radiation dose is 5 × 10⁴ Ci, the shield wall is 85 cm of heavy concrete, and two shielding windows are installed in the front [6]. Additionally, the maximum ventilation volume flow rate of the hot cell internal facility is kept at 4200 m³/h for a negative pressure, while the ventilation volume flow rate in the working space outside the hot cell facility is kept at 2722.5 m³/h [7].

Figure 8. 3D view of PIEF 9405 3.5 m

4.0 m 1.5 m

85 cm

22

2.10 Radioactive waste handling process in hot cell facility

2.10.1 Spent nuclear fuel processing in Korea

Among the radioactive wastes generated in Korea, the final management policy for SNF has been reserved. Therefore, the cumulative amount of SNF generated from NPP continues to increase.

Research on various methods such as direct disposal or reuse will be conducted in hot cell facilities.

Therefore, next-generation management processes for SNF, such as the extraction of spent fuel rods, cutting of fuel rods and the compression treatment process of structural waste, should be carried out in the hot cell facility. In Korea, technology development for the SNF process will be carried out at PIEF. In PIEF, SNF can be severed by remote control for destructive testing. It was confirmed that almost no dust was generated when spent fuel rods were cut in PIEF, and a minute amount of nuclear fuel powder was collected from the cut area during cutting [8]. In addition, it was confirmed that the cutting surface of the fuel rod maintained good circularity, but the cutting surface of the cladding material was slightly rough [8]. In the general oxide layer appearance of nuclear fuel rods investigated in commercial reactors, a black protective oxide layer could be observed up to a certain distance from the bottom of the fuel rod [8]. It was confirmed that this black oxide layer changed to a thick white oxide film. Through these experiments, it is advantageous in terms of aerosol generation during cutting and the risk of fire. Since the circularity is well maintained, it is expected to be appropriate for the spent fuel rod cutting method.

23

2.10.2 Development of radioactive waste process technology in abroad

In most countries that reprocess SNF, spent fuel is cut by shearing and decladding is performed by treating the cut spent fuel with a chemical solution such as nitric acid. The shearing method includes in the case of wet reprocessing, after disassembling the spent fuel assembly and cutting it in units of rods the first method is generally mainly used. In the Institute for Transuranium facility (ITU) in Germany, spent fuel rods transferred in rod units from an adjacent cell, which is a spent fuel assembly disassembly cell, are being wet cut using a diamond wheel cutter for post irradiation testing of spent fuel [8]. In addition, the Pilot conditioning plant (PKA) facility in Germany is a research and development facility for making spent nuclear fuel into a suitable form for final disposal. The main function of the PKA facility is to perform the spent fuel consolidation process, as well as dismantle the spent fuel assembly, separate the fuel rods, cut them in half, and pack them into dense canisters [8].In addition, in 2010, the Blue-Ribbon commission on America’s nuclear future (BRCANF) was established in the United States to seek new management alternatives for spent nuclear fuel. In 2013, a strategy for the management and disposal of SNF and high-level waste was established. The main contents of the management strategy were decided to operate a pilot interim storage facility in 2020 and a large-capacity interim storage facility in 2025. To this processing, Sandia National Laboratories/New Mexico (SNL/NM) in Albuquerque, New Mexico, manages and repackages all trans uranium (TRU)-related waste at auxiliary hot cell facility (AHCF) [9].

24

2.11 Radiation work and dose regulation measures

According to ICRP 30, the concentration in air is derived from the annual intake limit based on 2000 hours of work per year with light work of 50 weeks per year and 40 hours per week. Therefore, the annual working hours of radiation workers were assumed to be 2000 hours [10]. Derived air concentration (DAC) divides the annual limit intake (ALI) of radionuclides by 2,400 m³ (2,000 working hours per year, respiration rate 1.2 m³/h), which is the amount of air a worker inhales during one year of work, to determine the radiation concentration inhaled during annual work. According to the Nuclear Safety Commission Notification No. 2016-16, Article 7 Paragraph 2 of the Standard on Radiation Protection etc.,when the type of radioactive material is known and there are two or more radionuclides, the ALI is limited to the inhaled or oral intake amount where the sum of the ratios of each radionuclide to the inhalation or ALI of each radionuclide is 1.

Table 4. Dose Limit of Enforcement Decree in the Nuclear Safety Act [11]

Classification Radiation Worker Persons with Frequent Access and Persons Engaging in Transport

Public

Effective Dose Limit

100 mSv for five years within the scope not exceeding 50 mSv per

annum

12 mSv per annum 1 mSv per annum

Equivalent Dose Limit

Crystalline 150 mSv per annum 15 mSv per annum 15 mSv per annum

Hands, Feet

and Skin 500 mSv per annum 50 mSv per annum 50 mSv per annum

25

III. Materials and Methods

3.1 Design of hot cell facility

3.1.1 Auxiliary hot cell facility

The AHCF of the SNL, for packaging radioactive waste and materials for reuse, recycling or eventual disposal was utilized a hot cell facility. [12]. The AHCF, which has an internal volume of 3.08 x 3.08 x 3.64 m³, is depicted in Figure 9.The shielding wall of AHCF is a triple structure unlike the existing hot cell facility, and the inner, middle, and outer shielding structures are made of concrete, sand, and concrete, and the thickness is 20cm, 61cm, and 20cm, respectively. The floor height of the shield wall, excluding the inside of the AHCF, is 46cm from the floor. The inside floor height of the hot cell is 56 cm. The roof has three layers of protection: 20 cm of concrete, 30 cm of sand, and 20 cm of concrete.

Figure 9. 3D view of AHCF

The south and west walls of the AHCF are built with lead glass windows that are as thick as the shielding wall. Additionally, the highest value of ventilation volume flow rate of the internal facility is kept at 2,548 m³/h for a negative pressure, while the ventilation volume flow rate in the working space outside the AHCF is kept at 1,189 m³/h. The air change rate, which measures the number of exchanges of the amount of air flowing into the hot cell inside in relation to the volume of the building space with the AHCF, is 4.89-73.79/h. The internal volume of the hot cell is 34.53 m³, and its air change rate is that Inner barrier wall (concrete): 20 cm

Middle barrier wall (sand): 61 cm

Lead glass

Outer barrier wall (concrete): 20 cm

3.08 m 3.64 m

26

number of exchanges per hour. Additionally, the design standard strength was fixed at 0.22 g, which is consistent with the AHCF design base earthquake strength of 0.22 g, meaning that a stronger earthquake might demolish the structure. Also, work aera volume is 7395.9 m^3.

27

3.2 Decommissioning radioactive waste from PWR

The metallic decommissioning waste generated during the decommissioning of Kori Unit 1, a PWR type of NPP, is radioactive waste from the pressure vessel and internal construction of the reactor [13].

3.2.1 Reactor pressure vessel

Inside the reactor, steady nuclear reactions can take place in the cylindrical pressure vessel. A reactor pressure vessel's excellent safety and integrity are required because it cannot be changed during the lifetime of an NPP. Because the pressure vessel in the reactor is made of mild steel and is consecutively exposed to neutron radiation while the NPP is operating, the steel used for the pressure vessel needs to be strong and fracture-resistant [13]. The material for the reactor pressure vessel is made up of a core beltline metal with a size of 20 × 22 × 17 cm³ and a core weld metal with a size of 20 × 12 × 17 cm³ [13]. By separation and decontaminating the reactor structure, these samples are removed from the reactor and transported to a AHCF for storage and separation. It is possible that the composition of the reactor was altered by high-capacity neutrons during the operating life of the NPP for 38 years.

Therefore, it is intended to provide important information for estimating the radiological characteristics of the reactor pressure vessel of the PWR. [13].

3.2.2 Internal structure of the reactor

The reactor's core barrel, which is the inside structure, is built to stabilize the nuclear reaction in the fuel bundle and control rod assembly and transfer the various loads placed to the reactor pressure vessel.

This core barrel of internal structure in the reactor is made of carbon steel. The upper and lower portions of the vertical and girth welds were cut using the core barrel metal's material cutting technique [13].

The core barrel sample was also separated through cutting and disinfection, and then sent to the AHCF to undergo a further cutting operation. This process was like that used to separate the reactor pressure vessel. The core barrel sample of reactor internal structure had a volume of 10 × 20 × 4.445 cm³ [13].

28

Figure 10. Typical Pressurized Reactor

29

3.3 Nuclide inventory evaluation of decommissioning radioactive waste

Using the report as a source, such as NUREG/CR-3474 [14], The specifications and aspects of the radio source utilized to treat PWR dismantled waste at the AHCF were established. The procedure of cutting the metallic dismantled radioactive waste was carried out for both mild and stainless steel because most of the large metal structures that needed to be processed in the AHCF were built of SUS304 stainless steel and mild steel [14]. ⁶⁰Co, ⁵⁵Fe, ⁵⁴Mn, and ⁶³Ni are the main nuclides of SUS304 stainless steel. The nuclide with the highest radioactivity concentration right after closure was ⁵⁵Fe, while the largest share of radioactivity began to accumulate in ⁶⁰Co about 10 years after shutdown. Even though ⁶⁰Co isn’t a fundamental element of stainless steel, it does exist as an impurity and is created when ⁵⁹Co (n, γ) reacts with ⁶⁰Co [14]. 10 years after the PWR had been in service for 30 effective full power years (EFPY), six nuclides (⁶⁰Co, ⁶³Ni, ⁵⁵Fe, ³H, ⁵⁹Ni, and ¹⁴C) accounted for more than 99% of the nuclide inventory in the reactor structure. The ratio of these six major nuclide radioactivity concentrations in the core beltline of reactor pressure vessel is shown in Table 1. For the core barrel of reactor internal structure, Table 2 provides comparable data. Between the reactor pressure vessel and the core barrel, there was a comparable ratio of radioactivity concentration for the six nuclides. The external dose coefficient of ⁶⁰Co at 4.23E-01 nSv/h per Bq/m³ was identified as the highest compared to ³H: 9.05E-10, ¹⁴C: 6.29 E-06, ⁵⁵Fe: 1.94E-11, ⁵⁹Ni: 2.32E-06, and ⁶³Ni: 4.10E-08 nSv/h per Bq/m³ [14]. As a result, it is anticipated that measuring the radioactivity of ⁶⁰Co will enable the evaluation of radiological safety. The other 34 species of isotopes that compose up the core barrel of SUS304 stainless steel were also found to account for a very small fraction, with ¹³³Ba having the largest proportion (0.001832%) compared to the other nuclides [14]. The majority of other isotopes exhibit decay within 10 years, which results in a large reduction in quantity, or exhibit extremely low initial radioactivity concentrations, which make them inconsequential in terms of dose.

30

Table 5.Nuclide inventory of core beltline (reactor pressure vessel) [14]

Isotope Half-life

(yr)

Radioactivity Concentration

(Bq/g)

Radioactivity Portion

(%)

3H 12.3 2.74 × 10⁵ 0.4000

14C 5,730 2.70 × 10⁵ 0.0394

55Fe 2.7 1.68 × 10⁸ 24.4784

59Ni 80,000 1.60 × 10⁶ 0.2324

60Co 5.272 3.18 × 10⁸ 46.4526

63Ni 100 1.97 × 10⁸ 28.7494

Table 6. Nuclide inventory of core barrel (reactor internal structure)[14]

Isotope Half-life

(yr)

Radioactivity Concentration

(Bq/g)

Radioactivity Portion

(%)

3H 12.3 2.02 × 10⁵ 0.07

14C 5,730 1.11 × 10⁵ 0.04

55Fe 2.7 6.82 × 10⁷ 23.41

59Ni 80,000 7.03 × 10⁵ 0.24

60Co 5.272 1.39 × 10⁸ 47.78

63Ni 100 8.29 × 10⁷ 28.46

31

3.4 Cutting scenario of radioactive waste

The AHCF processed samples taken from the reactor pressure vessel and core barrel. It is possible to do an examination of the actual reactor pressure vessel's irradiation embrittlement using the processed core beltline metal, as well as to get crucial information for preserving the integrity of the reactor's long- term operation [14]. It is necessary to cut the reactor pressure vessel's core beltline metal for this reason.

When a cutting device was utilized for radioactive waste material in the AHCF, the emission of aerosols was assessed at this time to satisfy the emission management requirement, which is a radiation protection standard. A circular saw or a band saw was utilized for the cutting procedure. It was estimated that the band saw blade was 8 cm wide, 0.16 cm thick, and had a kerf thickness of 0.2 cm, while the circular saw blade had a diameter of 53 cm, a thickness of 0.38 cm, and a kerf thickness of 4 mm. The circular saw's thickness is double that of the band saw, as can be shown in Table 9, which results in a doubling of the volume of cutting loss.

3.3.1 Cutting scenario of reactor pressure vessel

Calculations were made to determine the volume of the sample and the cutting loss volume after the core beltline metal was uniformly cut three times to a length of 22 cm in the reactor pressure vessel.

The core weld metal was divided into four processed sample pieces, each of which was uniformly sliced three times to a length of 20 cm.

3.3.2 Cutting scenario of internal structure in reactor

Four processed sample pieces were created by uniformly cutting the reactor's core barrel three times to a length of 20 cm. For the core beltline and core barrel samples, Table 3 indicates the cutting loss volume while using circular and band saws.

32

Table 7.Data on decommissioning waste sample volume, amputated sample volume, and cutting loss volume

Decommissioing

Waste Material Sampling Location

Sample Density (g/cm

³

)

Sample Volume (cm

³

)

Circular saw Band saw

Amputated Sample Volume (cm

³

)

Cutting Loss Volume

(cm

³

)

Amputated Sample Volume

(cm

³

)

Cutting Loss Volume

(cm

³

) Reactor Pressure

Vessel Core beltline 7.82 20 × 22

× 17

20 × 5.2

× 17 408 20 × 5.35

× 17 204

Reactor Pressure

Vessel Core weld 7.82 20 × 12

× 17

4.7 × 12

× 17 244.8 4.85 × 12

× 17 122.4

Reactor Internal Structure

Core barrel

7.94 10 × 20

× 4.4

10 × 4.7

× 4.4 53.34 10 × 4.85

× 4.4 26.67

33

3.5 Aerosol production during cutting process

Aerosols are created during mechanical cutting due to friction and heat energy [16]. Particles are suspended because of the saw blade's motion, which causes heat energy to be produced from the friction between the cutting material and the saw blade. Physical elements and the saw blade's speed like the length of the contact surface and the horizontal drag When the saw blade touches the material are what determine these variables, such as the friction force and thermal energy. During cutting with radioactive metals in AHCF, the airborne contaminant term resulting from mechanical cleavage was evaluated. This term is strongly related to the internal dose of worker. Coarse, fine, and ultrafine particles are created during the cutting process and discharged into the atmosphere [16]. In comparison to other activities, mechanical cleavage produces high mass concentrations of airborne particles and airborne source terms.

During the procedure, 97% of the particles produced are ultrafine particles. With an increase in produced particles, the particle mass concentration increases. Aerosols with particle lengths of 5–560 nm were identified using reactive differential mobility spectroscopy [17]. During the cutting operation, a value of 554.1 nm for the determined activity median aerodynamic diameter was validated [17].

34

3.6 Aerosol mass

The ventilation system of hot cell facility kept the aerosol's radioactivity concentration constant throughout the cutting process. The radioactivity concentration from the aerosol generated in the cleaving process was calculated from the aerosol mass, as given in Eq. (1):

𝐴_𝑚= 𝐿_𝑣𝑜𝑙× 𝜌_𝐴× 𝐴𝑅𝐹 (1)

where 𝐴_𝑚 is the aerosol mass (g), 𝐿_𝑣𝑜𝑙 is the cutting loss volume (cm³), 𝜌_𝐴 is the density (g/cm³), and 𝐴𝑅𝐹 is the aerosol release fraction (%).

A reciprocating saw generates 0.0149% of the ARF when cutting a piece of 5 cm mild steel since the loss of mass per cutting length is 850.4 g/m and the quantity of aerosol production is 0.127 g/m [18].

Additionally, 0.0671% of the ARF is produced when using a reciprocating saw to cut a SUS304 stainless steel sample because of the 1936 g/m loss mass per unit cutting length and the 1.30 g/m aerosol formation amount. The reciprocating saw's cutting force on the workpiece is 50 N, and its practical power rating is roughly 500 W. To guarantee measurement reproducibility, these experiments were conducted again [18]. These conditions were followed when cutting in this study.

3.6.1 Aerosol mass of reactor pressure vessel

Using Eq. (1), It was found that when the core beltline metal was cut with a circular saw, the aerosol's mass was 0.475 g, and when it was cut with a band saw, the aerosol's mass was 0.238 g. Moreover, the aerosol's mass was 0.143 g when the core welding part was cut using a band saw as opposed to 0.285 g when it was cut with a circular saw.

3.6.2 Aerosol mass of internal structure in reactor

Using Eq. (1), the mass of the aerosol was 0.284 g when the core barrel metal was cut with a circular saw and 0.142 g when it was cut with a band saw.

35

3.7 Main routes of exposure for radiation workers

3.7.1 Normal operation of hot cell facility

The exposure of radiation workers when working in the operating area is divided into internal exposure and external exposure. When the ventilation system was operating with leakage and dose was evaluated based on a leak rate of 1 %/hr under the assumption of a permanent leak in the air in the operator's space according to ISO 10648-2 Class3 (Condition of a Static Pressure Aseptic Circuit Breaker) under normal operating conditions [19].

3.7.2 Fire Accidental scenario of hot cell facility

The workers' exposure dose was assessed with the assumption that the ventilation system would be turned off in the event of an unintentional fire.

3.7.3 Earthquake scenario of hot cell facility

The dose to which employees were exposed was calculated assuming the gap between the lead glass and the shielding wall is 2 cm. If an earthquake occurred that was stronger than depending on the design criteria, the ventilation system may malfunction.

36

3.8 Airborne source term concentration

The airborne source term concentration in the AHCF is calculated using Eq. (2) using the aerosol mass in Eq. (1).

𝐴

_𝑠𝑡

=

^{𝑅𝑠 × 𝐴𝑚}

𝑉

(2)

Where in a hot cell facility, 𝐴_𝑠𝑡 is the airborne source term concentration (Bq/m³), 𝑅_𝑠 is the sample's radioactivity content (Bq/g), and V is the hot cell facility's interior space volume (m³).

37

3.9 Indoor particle concentration

A mass-balance model Eq. (3) was used to calculate the concentration of aerosol radioactivity while accounting for ventilation, considering the generated amount and ventilation volume flow rate of the source [20]. Under the premise that aerosols were not released under the working parameters of 40 hours per week and 2,548 m³/h in the AHCF, Eq. (3) was used to calculate the indoor particle change concentration.

^{𝑑𝐶𝑖𝑛}

𝑑𝑡 = 𝑃𝜆𝐶_𝑜𝑢𝑡 + (1 − 𝜂_𝑣)^𝑄

𝑉𝐶_𝑜𝑢𝑡 +^𝐺

𝑉− 𝜆𝐶_𝑖𝑛−^𝑄

𝑉𝐶_𝑖𝑛− 𝑘𝐶_𝑖𝑛− 𝜂_𝑓^𝑄𝑓

𝑉 𝐶_𝑖𝑛

(3)

Where 𝐶_𝑖𝑛 is the indoor particle concentration (Bq/m³), 𝐶_𝑜𝑢𝑡 is the indoor particle concentration (Bq/m³), 𝜆 is infiltration/exfiltration rate (h⁻¹), P is penetration coefficient (-), 𝜂_𝑣 is particle removal efficiency of ventilation system (-), 𝜂_𝑓 is particle removal efficiency of filtration system (−), 𝑄 is the ventilation volume flow rate (m³/h), 𝑄_𝑓 is the filtration flow rate (m³/h), 𝐺 is the indoor generation

rate (Bq/h), 𝑘 is deposition rate (h⁻¹ ), 𝑉 is internal volume(m³) and 𝑡 is the processing time (h) [19].

^{𝑑𝐶𝑖𝑛} 𝑑𝑡

=

^𝐺

𝑉

−

^𝑄

𝑉

𝐶

_𝑖𝑛 (4)

Eq. (4) was derived under the assumption that there is no leakage or inflow from the working area among the particles.

𝐶

_𝑖𝑛

=

^𝐺

𝑄

[1 − exp (−

^𝑄

𝑉

𝑡)]

(5)

As a result of solving the differential equation using Eq. (4), Eq. (5) was finally derived. Where 𝑡 value is the 40 hours, 𝐺 is the indoor generation rate (Bq/h), and 𝑄 value is the 2548 m³/h.

38

3.10 MCNP evaluation methodology

MCNP was used to assess external exposure, and dose evaluation was carried out under the assumption that the workers were at work. A location 10 cm away from the AHCF was chosen to evaluate the exposure dose rate. For evaluating dose rates, the dose to the personnel was assessed during ordinary operation and during an accident when the ventilation system was used. When the ventilation system was operating, the exposure dose was calculated, and aerosol radioactivity concentrations reached equilibrium with no leaks when the ventilation system was operating. It was assumed that the shielding wall wasn't leaking during regular operation. The two situations were divided into two categories and assessed in the case of an accident. First, the worker’s exposure dose was calculated by contrasting the normal operation and accident scenarios in Figure. 11 and assuming that the ventilation system would be turned off in the event of an accidental fire. It was estimated that there was no leakage from the shielding wall, as is normal. Second, the ventilation system would malfunction if an earthquake occurred that was stronger than anticipated. The dose to which the workers were exposed was calculated based on a 2 cm gap between the lead glass and shielding wall. In addition, gamma nuclides (⁶⁰Co, ⁵⁵Fe,

59Ni) were evaluated with tally6 and beta nuclides (³H, ¹⁴C, ⁶³Ni) with tally4, considering the airborne source term and indoor particle concentration. The radiation-controlled area's weekly external radiation exposure rate is 400 µSv, as stated in Article 3 of the technical guidelines for radiation safety management. A weekly workweek of 40 hours is implemented when the dose limit value is 0.01 mSv per hour.

Figure 11. Geometry of considering ventilation and non-ventilation in AHCF

39

IV. Results and discussions

4.1 Environment radiation emission evaluation

With an effectiveness of 99.97%, the high efficiency particulate air (HEPA) filter expelled the airborne source term from ventilation to the outside. As a result, 0.03% of the interior particle concentration was carried outside. Tables 8~13, which divide the hot cell facility's equipment into situations with and without ventilation, display the amount of radioactivity in aerosols in each category.

When the core beltline, core weld, and core barrel metals were cut with circular and band saws while taking ventilation into consideration, the concentration of the airborne source term is shown in Figures 12–14.