The Eleventh International Symposium on
Radiation Safety and Detection Technology

Abstracts

4-7 July 2023,

Hanyang University, Seoul, Korea

Contents
Invitation

3

Conference Organization

4

Sponsored & Supporting Organization

6

Program at a Glance

7

Program Summary

8

Program Detailed

10

Monday, July 3
Tuesday, July 4
Wednesday, July 5
Thursday, July 6
Friday, July 7 (PHITS Tutorial)
Abstracts Summary
Invited_speakers

40

Oral 1. Radiation Transport and Shielding

62

Oral 2. Radiation Dosimetry

76

Oral 3. Radiation Detection and Sensor Technology

108

Oral 4. Environmental Radiation Measurement and Assessment

165

Oral 5. Radiological Risk Management

187

Oral 6. Education, Training and Policy in Radiation Safety

198

Oral 7. Radiological Emergency Planning and Preparedness

202

Oral 8. Radioactive Waste and Current Radiological Issues

209

Poster 1. Radiation Transport and Shielding

252

Poster 2. Radiation Dosimetry

272

Poster 3. Radiation Detection and Sensor Technology

312

Poster 4. Environmental Radiation Measurement and Assessment

367

Poster 5. Radiological Risk Management

413

Poster 6. Education, Training and Policy in Radiation Safety

426

Poster 7. Radiological Emergency Planning and Preparedness

435

Poster 8. Radioactive Waste and Current Radiological Issues

456

Welcome Remark
I warmly welcome all of you who are participating in the 11th International Symposium on
Radiation Safety and Detection Technology (ISORD-11).
The aim of this symposium is to provide a dynamic platform where researchers, professionals,
and experts can come together to share and discuss a wide range of research topics pertaining
to radiation safety and detection technology. I hope that this symposium will be a very
productive and enlightening experience for all of you.
Initiated in 2001 through a joint effort by Hanyang University, Tohoku University, and the
China Institute for Radiation Protection, the ISORD Symposium was designed to offer
an international platform for participants to exchange ideas, insights, and information onradiation safety and detection technology, as well as addressing general issues concerning
radiation protection.
Over the past two decades, the ISORD Symposium has played a very important role inadvancing radiation safety and detection technology in Korea, China, and Japan, providing
numerous opportunities for scientific and technological progress, as well as promoting
invaluable information exchange.
During this symposium, researchers and experts from scientific, technological, and industrial
communities engaged in radiation safety and detection technology come together to present
and discuss their most recent research findings. I encourage all of you to actively participate
in these discussions and fully embrace this unique learning opportunity.
In addition, I hope that this symposium will facilitate the strengthening of partnerships andthe expansion of cooperation, which are the primary objectives of hosting this event.
Thank you,

Chan Hyeong Kim, Ph.D.
Chair
ISORD-11 Organizing Committee

Conference Organization
Organizer: ‌Innovative Technology Center for Radiation Safety (iTRS)
Korean Association for Radiation Protection

Monitoring and Steering Committee• Chair – Jong Kyung KIM (Hanyang University)• Secretary – Chang Ho SHIN (iTRS, Hanyang University)• Chan Hyeong KIM (Hanyang University)• Kyo-Youn KIM (Korea Atomic Energy Research Institute)• Hee-Seock LEE (Pohang Accelerator Laboratory/POSTECH)• Huating YANG (Chinese Society of Radiation Protection)• Senlin LIU (China Institute of Atomic Energy)• Liye LIU (China Institute for Radiation Protection)• Qun LIU (China Institute for Radiation Protection)• Rui QIU (Tsinghua University)• Takashi NAKAMURA (Tohoku University)• Mamoru BABA (Tohoku University)• Kenji ISHIBASHI (Kyushu University)• Takatoshi HATTORI (Central Research Institute of Electric Power Industry)• Tetsuo IGUCHI (Nagoya University)• Hiromi YAMAZAWA (Nagoya University)• Hiroyuki TAKAHASHI (University of Tokyo)

Honorary Co-chairs• Takashi NAKAMURA (Tohoku University, Japan)• Jong Kyung KIM (Hanyang University, Korea)• Qun LIU (China Institute for Radiation Protection, China)

Organizing Committee• Chair – Chan Hyeong KIM (Hanyang University)• Vice Chair – Hee-Seock LEE (Pohang Accelerator Laboratory/POSTECH)• Secretary General – Yeon Soo YEOM (Yonsei University)• Secretary General Assistant – ‌Bangho SHIN (Hanyang University),
Seung Hyun LEE (Hyundai Engineering)• Program Chair – Tae Young KONG (Chosun University)• Publication Chair – Geehyun KIM (Seoul National University)4• Treasurer – Chang Ho SHIN (Hanyang University)• Web Master – Min Cheol HAN (Yonsei University College of Medicine)• Student Competition Chair – Wi-Ho HA (Korea Atomic Energy Research Institute)• Sponsorship Chair – Tai Jin PARK (Korean Association for Radiation Application)

Promotion & Editorial Committee[ Korean Members ]• Yoonsun CHUNG (Hanyang University)• Kwang Pyo KIM (Kyung Hee University)• Hee Reyoung KIM (Ulsan National Institute of Science and Technology)• Wonho LEE (Korea University)• Jungil LEE (Korea Atomic Energy Research Institute)• Jin-Woo LEE (Korea Atomic Energy Research Institute)• Jeong-In KIM (Korea Hydro & Nuclear Power)• Jaeryong YOO (Korea Institute of Radiological & Medical Sciences)• Seung-Young JEONG (Korea Institute of Nuclear Safety)• Keon Wook KANG (Seoul National University College of Medicine)[ Chinese Members ]• Senlin LIU (China Institute of Atomic Energy)• Qun LIU (China Institute for Radiation Protection)• Liye LIU (China Institute for Radiation Protection)• Rui QIU (Tsinghua University)• Huating YANG (Chinese Society of Radiation Protection)• Ling CHEN (China Institute of Atomic Energy)• Xiaobin XIA (Shanghai Institute of Applied Physics, Chinese Academy of Sciences)• Chuan WANG (CNNC Nuclear Power Operations Management Limited)• ‌Quanfu SUN (National Institute for Radiological Protection, Chinese Center for Disease
Control and Prevention)[ Japanese Members ]• Hiroshi NAKASHIMA (Hokkaido University)• Hiroyuki TAKAHASHI (University of Tokyo)• Hiroshi YASUDA (Hiroshima University)• Tatsuo TORII (Fukushima University)• Keisuke MAEHATA (Teikyo University)• Hideki TOMITA (Nagoya University)• Shunsuke KUROSAWA (Tohoku University)• Yukio SAKAMOTO (ATOX Co., Ltd.)
5

Sponsored Organization▶

Korea Nuclear International Cooperation Foundation (KONICOF), Korea▶ Seoul Metropolitan Government, Korea▶ RADCORE, Co., Ltd., Korea▶ Korea Reactor Integrity Surveillance Technology (KRIST), Korea

Supporting Organization▶

Japan Health Physics Society (JHPS), Japan▶ Atomic Energy Society of Japan (AESJ), Japan▶ Tohoku University, Japan▶ China Society of Radiation Protection (CSRP), China▶ China Institute for Radiation Protection (CIRP), China▶ China Institute of Atomic Energy (CIAE), China▶

Korean Nuclear Society (KNS), Korea

6

Program at a GlanceTime

08:00 - 08:30

3 July (Mon)

4 July (Tue)Registration

09:00 - 09:30

ICRP Present andFuture: Kun-Woo Cho
and Christopher H.
Clement(Room 612)

09:30 - 10:00

Invited Talk: Hiroyuki
Takahashi(Room 612)

10:00 - 10:30
10:30 - 11:00

11:30 - 12:00

Poster Display(Room 608)

Poster Display(Room 608)

Break Time

12:00 - 13:30

Break Time

Lunch

13:30 - 14:00
14:00 - 14:30

Scientific Session(Room 612)&
Student Competition(Room 604, 606)

15:00 - 15:30

Break Time

15:30 - 16:00

Scientific Session(Room 612)&
Student Competition(Room 604, 606)

14:30 - 15:00

16:00 - 16:30
16:30 - 17:00

PosterPresentation(Room 608)

Scientific Session(Room 604, 606, 612)

Scientific Session(Room 604, 606, 612)

PHITS Tutorial(Room 606)

Closing Session(Room 612)

PosterPresentation(Room 608)
Technical Tour(KIRAMS)

Break Time
Poster Display(Room 608)

Registration

Scientific Session(Room 604, 606, 612)

Scientific Session(Room 604, 606, 612)

Scientific Session(Room 604, 606, 612)

Registration

7 July (Fri)

Invited Talk: Jan
Schuemann, Nina
Petoussi-Henss, andKeon Wook Kang
(Room 612)

Invited Talk: Liye Liu,
Tatsuhiko Sato, andRui Qiu(Room 612)

Break Time

Registration

6 July (Thur)

Registration

Opening Session(Room 612)

08:30 - 09:00

11:00 - 11:30

5 July (Wed)Poster Display(Room 608)

PHITS Tutorial(Room 606)

17:00 - 17:30
17:30 - 18:00

EveningTime

Welcome Reception(18:00 - 19:00)

Conference Dinner(18:00 - 21:00)- Award Ceremony (612)- ‌Korean Traditional
Performance (612)- Dinner (6F, Restaurant)

MSC DinnerMeeting(19:00 - 21:00)

Presentation Schedule
ICRP Present
and Future

Invited Talk

Invited Speakers

Affiliation

4 July, 09:00

Kun-Woo Cho

Korea Institute of Nuclear Safety (KINS)

4 July, 09:30

Christopher H. Clement

4 July, 10:00

Hiroyuki Takahashi

5 July, 08:30

Liye Liu

5 July, 09:00

Tatsuhiko Sato

5 July, 09:30

Rui Qiu

6 July, 08:30

Jan Schuemann

6 July, 09:00

Nina Petoussi-Henss

Federal Office for Radiation Protection (BfS)

6 July, 09:30

Keon Wook Kang

Seoul National Unversity College of Medicine

International Commission on Radiological Protection (ICRP)The Unversity of Tokyo
China Institute for Radiation Protection (CIRP)
Japan Atomic Energy Agency (JAEA)
Tsinghua University

7

Massachusetts General Hospital and Havard Medical School

Program Summary
Monday, July 3
Time

Room 612

Room 606

08:30 – 17:00

Registration (Lobby)

18:00 – 19:00

Welcome Reception (Lobby)

19:00 – 21:00

MSC Dinner Meeting (Room 606)

Room 604

Room 608

Room 604

Room 608

Tuesday, July 4
Time

Room 612

08:00 – 08:30

Registration (Lobby)

08:30 – 09:00

Opening Session (Room 612)

09:00 – 10:00

ICRP Present and Future (Room 612)

10:00 – 10:30

Invited Talk (Room 612)

10:30 – 11:00

Coffee Break

11:00 – 12:00

Session O3

12:00 – 13:30

Lunch
Session O2

13:30 – 15:00
15:00 – 15:30

Room 606

Poster Display

Session O7

Session O2

Student
Competition:
Session O3

Student
Competition:
Session O1
Session O2

Session P1
Session P2
Session P3
Session P4

Student
Competition:
Session O3
Session O5
Session O6
Session O8

Student
Competition:
Session O2
Session O4

Poster Display

Room 606

Room 604

Coffee BreakSession O2

15:30 – 17:00

Wednesday, July 5
Time

Room 612

08:00 – 08:30

Registration (Lobby)

08:30 – 10:00

Invited Talk (Room 612)

10:00 – 10:30

Coffee Break

10:30 – 12:00

Session O3

12:00 – 13:30

Lunch
Session O3

13:30 – 15:00

Session O1

Session O8

Session O1
Session O4

Session O8

8Room 608

Poster Display

Session P4
Session P5
Session P6
Session P7
Session P8

Time
15:00 – 15:30
15:30 – 17:00
18:00 – 21:00

Room 612

Room 606

Room 604

Room 608

Coffee BreakSession O3

Session O4

Session O8
Session O5

Poster Display

Conference DinnerThursday, July 6
Time

Room 612

08:00 – 08:30

Registration (Lobby)

08:30 – 10:00

Invited Talk (Room 612)

10:00 – 10:30

Coffee Break

10:30 – 11:30

Room 606

Session O3

Session O4

11:30 – 12:00

Closing Session (Room 612)

12:00 – 13:30

Lunch

13:30 – 17:00

Technical Tour (KIRAMS)

Room 604

Session O5
Session O6

Friday, July 7
Time

Room 612

Room 606

08:30 – 12:00

PHITS Tutorial (Room 606)

12:00 – 13:30

Lunch

13:30 – 18:00

PHITS Tutorial (Room 606)

Session O1 / P1: Radiation Transport and Shielding
Session O2 / P2: Radiation Dosimetry
Session O3 / P3: Radiation Detection and Sensor Technology
Session O4 / P4: Environmental Radiation Measurement and Assessment
Session O5 / P5: Radiological Risk Management
Session O6 / P6: Education, Training and Policy in Radiation Safety
Session O7 / P7: Radiological Emergency Planning and Preparedness
Session O8 / P8: Radioactive Waste and Current Radiological Issues

9

Room 604

Program
Detailed
ProgramMonday, July 3
08:30-17:00
18:00-19:00

Registration
Welcome Reception

Room 606
19:00-21:00

MSC Dinner Meeting

Tuesday, July 4
08:00-08:30

Registration

Room 612
08:30-09:00
1.

Opening Session

Opening Address- Chan Hyeong Kim (Chair of ISORD11)2.

Welcome Address- Ki Jeong Lee (President of Hanyang University)

3.

Congratulatory Address 1- Jong Kyung Kim (Chair of ISORD Monitoring and Steering Committee & Honorary Co-chairs)

4.

Congratulatory Address 2- Sung Hwan Kim (President of Korean Association for Radiation Protection, KARP)

09:00-10:00

ICRP Present and Future

Moderator:

Jai Ki Lee, Tatsuhiko Sato

1.

Current Activities of the International Commission on Radiological Protection
Kun Woo Cho, Korea Institute of Nuclear Safety, Korea

2.

Review and Revision of the System of Radiological Protection: RP for the Next Generation
Christopher H. Clement, International Commission on Radiological Protection1010:00-10:30

Invited Talk

Moderator:

Jai Ki Lee, Tatsuhiko Sato

3.

Fast high granular optical Transition Edge Sensor with parallel biasing scheme forquantum applications
Hiroyuki Takahashi, The University of Tokyo, Japan

10:30-11:00

Coffee Break

11:00-12:00

Session O3

Chair:

Geehyun Kim, Hengguan Yi, Shunsuke Kurosawa

O3-01

Development of Two Types of Lightweight Portable Neutron Survey Meter
Takashi Nakamura et al., Fuji Electric Co. Ltd., Japan

O3-02

Detection of gamma-ray photons with halide lead perovskite single crystals for high energy
resolution
Xin Wang et al., School of Electronic Science and Engineering, Joint International ResearchLaboratory of Information Display and Visualization, Southeast University, China

O3-03

Low-noise and High sensitivity direct-conversion X-ray detectors based on perovskite
single-crystal heterojunctions
Yuzhu Pan et al., Joint International Research Laboratory of Information Display and
Visualization, School of Electronic Science and Engineering, Southeast University, China

O3-04

Application of Unfolding code MEALU to Neutron Spectrum Measurement using Multicylindrical Bonner Counter
Tetsuo Iguchi et al., Nagoya University, Japan

12:00-13:30

Lunch

13:30-15:00

Session O2

Chair:

Jungil Lee, Qinjian Cao, So Kamada

O2-05

Preliminary Phantom and Monte Carlo Space Radiation Dosimetric Studies for Lunar
Missions
Zirui Ye et al., School of Nuclear Science and Technology, University of Science and Technology
of China, China

O2-06

A script-based automatic commissioning method of virtual source model for independent
dose verification module
Bo Cheng et al., School of Nuclear Science and Technology, University of Science andTechnology of China, China|11

O2-07Age-dependent Iodine-131 S values Based on ICRP Reference Pediatric Phantoms
Soo Min Lee et al., Department of Radiation Convergence Engineering, Yonsei University,
Korea

O2-08

CPU-GPU Coupling Monte Carlo Acceleration Method for Dose Calculation of Mesh-type
Phantom
Shuchang Yan et al., Department of Engineering Physics, Tsinghua University, China

O2-09

Personalized

Cumulated

Activity

Distribution

Calculation

using

Biokinetic

Compartmental Model in PET Imaging
He Yutao et al., School of Nuclear Science and Technology, University of Science andTechnology of China, China

15:00-15:30

Coffee Break

15:30-17:00

Session O2

Chair:Han Ki Jang, Qinjian Cao, Takashi Nakamura

O2-13

Patient-Specific Dose Assessment for Nuclear Medicine Therapy Based on 177Lu-PSMA
Ziyi Hu et al., Department of Engineering Physics, Tsinghua University, China

O2-14

Establishment and first application of 4D dose calculation in China
Zhaoxing Liu et al., Department of Engineering Physics, Tsinghua University, China

O2-15

Establishment of High Dose Rate Pulsed γ Radiation Facility
Zhihui Tang, China Institute for Radiation Protection, China

O2-16

Establishment of radiation biological dose estimation curve based on automatic analysis ofdicentric chromosomes
Li Shufang et al., Department of Radiology and Environmental Medicine, China Institute for
Radiation Protection, China

O2-17

A performance study of a new personal neutron dosimeter based on SSNTD
Yan Jiao et al., China Institute for Radiation Protection, China

Room 606
11:00-12:00

Session O7

Chair:

Jaeryong Yoo, Junfang Zhang

O7-01

Validation and application of migration diffusion mode in nuclear accident consequence
evaluation system
Minghua Lyu et al., Department of Nuclear environmental science research, China Institute for
Radiation Protection, China

12O7-02

Investigation on Decision-Making Criteria for Public Protective Actions in Nuclear
Emergencies
Wi Ho Ha et al., Department of Nuclear Emergency Preparedness, Korea Atomic Energy
Research Institute, Korea

O7-03

A visual-based positioning and guidance approach for nuclear emergency response
manipulator
Qi Lv et al., China Institute for Radiation Protection, China

O7-04

Machine Learning-based Denoising Technique for Gamma Emission Tomography toInspect Partial Defects of Spent Nuclear Fuel
Hyung Joo Choi et al., Department of Radiation Convergence Engineering, Yonsei University,
Korea

12:00-13:30

Lunch

13:30-15:00

Student Competition: Session O3

Chair:

Jeongin Kim, Rui Qiu, Kenji Shimazoe

O3-05

Development of an in-house PMT voltage divider for the TDCR counting system
Agung Agusbudiman et al., Department of Precision Measurement, University of Science andTechnology (UST), Korea

O3-06

Study on FPD for high-energy X-ray CT for fuel debris screening
Akira Mikami et al., University of Tokyo, Japan

O3-07

Feasibility Verification of Deep Learning Based 4π Field of View Collimator-less Imaging
System Using Voxelated Single Volume GAGG(Ce) scintillator: A Monte-Carlo simulation
Ajin Jo et al., Health Science Research Center, Korea University, Korea

O3-08

Demonstration of Gamma Source Finding in the Field based on Fusion Data between 4π
Gamma Imaging and Simultaneous Localization and Mapping
Minato Kanda et al., Graduate school of Engineering, Nagoya University, Japan

O3-09

Research on Detection Efficiency Calibration by GEANT4 Simulation for the HPGe
Detector in HTR-PM
Rui Nie et al., Institute of Nuclear and New Energy Technology, Collaborative Innovation
Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor
Engineering and Safety of Ministry of Education, Tsinghua University, China

15:00-15:30

Coffee Break

15:30-17:00

Student Competition: Session O3 & O5 & O6 & O8

13

Chair:Jeongin Kim, Rui Qiu, Kenji Shimazoe

O3-10

High-energy photon counting based on Se-doped MAPbI3 single crystals
Yubing Xu et al., School of Electronic Science and Engineering, Joint International ResearchLaboratory of Information Display and Visualization, Southeast University, China

O5-01

Radiological Assessment of Loss of Coolant Accident from Shin Kori Nuclear Power Plant
Moses Oboo et al., Kepco International Nuclear Graduate School (KINGS), Department of
Nuclear Power Plant Engineering, Korea

O6-01

Tutorial Development for Digital Twin Simulation in Scenario-based Linear Accelerator
Daily QA Training
Wonhyeong Lee et al., Department of Nuclear Engineering, Hanyang University, Korea

O8-01

Factor Analysis of In-Situ Object Counting System(ISOCS) for Uranium Quantificationin Radioactive Waste Drums
Jihyun Ahn et al., School of Jeonbuk National University, Korea

O8-02

Development of AI-based Anomaly Detection Model for Nuclear Waste Data
Jinhyung Kwon et al., Korea Atomic Energy Research Institute (KAERI), Korea

Room 604
11:00-12:00

Session O2

Chair:Jaekook Lee, Rui Qiu, So Kamada

O2-01

Experimental evaluation of inter-fractional dosimetric changes in vaginal cuff
brachytherapy using in-house designed PMMA-agar phantom
Saerom Sung et al., Department of Radiation Convergence Engineering, Yonsei University,
Korea

O2-02

Impact of mesh-type Korean reference phantoms on dose calculations for computed
tomography (CT) scans
Yum i Lee et al., Department of Radiation Convergence Engineering, School of Yonsei
University, Korea

O2-03

Geometric Considerations in L-band EPR Tooth Dosimetry Using Radiation Treatment
Plans
Jong In Park et al., Ionizing Metrology Group, Korea Research Institute of Standards andScience, Korea

O2-04

GPU-Based Fast Monte Carlo Code for Carbon Ion Therapy
Jin Wu et al., School of Nuclear Science and Technology, University of Science and Technology
of China, China

14

12:00-13:30

Lunch

13:30-15:00

Student Competition: Session O1 & O2

Chair:

Wi Ho Ha, Junfang Zhang, Shunsuke Kurosawa

O1-01

Evaluation of Shielding Material against Galactic Cosmic Radiation: Monte Carlo
Simulations with ICRP-110 Reference Phantom
Li Xiaoyu et al., School of Nuclear Science and Technology, University of Science andTechnology of China, China

O1-02

Research on Complex Source Term Inversion and Radiation Field Reconstruction Method
Based on Deep Neural Networks
Yisheng Hao et al., Department of Engineering Physics, Tsinghua University, China

O2-10

Analysis of Occupational doses in Korean nuclear power plants over the past ten years
Chang Ju Song et al., School of Chosun University, Korea

O2-11

Development of Virtual Dose Monitoring System for Interventional Radiology (VIR)
Gahee Son et al., Department of Nuclear Engineering, College of Engineering, Hanyang
University, Korea

O2-12

Dose coefficients for soil contaminations produced using mesh-type Korean reference
phantoms (MRKPs)Ji Won Choi et al., Department of Radiation Convergence Engineering, Yonsei University,
Korea

15:00-15:30

Coffee Break

15:30-17:00

Student Competition: Session O2 & O4

Chair:

Wi Ho Ha, Junfang Zhang, Shunsuke Kurosawa

O2-18

Fast and Accurate GPU-Based Monte Carol Independent Dose Verification Platform forProton Therapy
Shijun Li et al., School of Nuclear Science and Technology, University of Science andTechnology of China, China

O2-19

Demonstration of Realtime Monte Carlo Person-specific Organ Dose Computing involving
Automatic Segmentation of 10,000 Patients
Zirui Ye et al., School of Nuclear Science and Technology, University of Science and Technology
of China, China

O4-01

Characterization of dose- and time-dependent radiation effects on human skin bymultiomics and histological analyses
JiEung Kim, et al., Department of Nuclear and Quantum Engineering, Korea Advanced Institute15

of Science and Technology (KAIST), Korea
O4-02

Development of a measuring method of cosmic-ray muon momentum distribution using
drift chambers
Naoto Nakagami et al., Kyushu University, Japan

O4-03

Determination of Ra-226 in Biological Samples via α-spectrometry Using Ra-223 as Yield
TracerPing Xu et al., University of Science and Technology of China, China

Room 608
Poster Session
08:30-12:00

Poster Display

12:00-13:30
13:30-15:00

Lunch
Poster Presentation

15:00-17:00

Poster Display

Session P1
P1-01

Operation test for the muon beam transport system of the μSR facility at RAON using abeta radiation source
Kihong Pak et al., Department of Nuclear Engineering, Hanyang University, Korea

P1-02

Validation of Nuclear Data Libraries Using Radiation Shielding Benchmarks in SINBAD
Haelee Hyun et al., Korea Atomic Energy Research Institute, Korea

P1-03

Study of transformation of HTO in lettuce for atsmospheric release of tritiated water
Qiao Cao et al., Institute of Nuclear Environmental Science, China Institute for Radiation
Protection, China

P1-04

Assessment of Dose Calculation Algorithms by Monte Carlo Method(MCNP)
Changyeon Yoon, KHNP Central Research Institute, Korea

P1-05

Analysis of Dose Assessment Methodology by Boundary Voxel of Irregular Shape Source
Based on Point Kernel Method
Min Seong Kim et al., KyungHee University, Korea

P1-06

Evaluation of Neutron effect for multi slit collimator in proton mini beam radiation
therapy
Chanil Jeon et al., Department of Health Sciences and Technology, SAIHST, Sungkyunkwan
University, Korea

P1-07

Research on 3D Radiation Field Calculation Based on CAD Modeling
Zeyu Long et al., China Institute for Radiation Protection, China

16

P1-08

Radiation Shielding and Activation Analyses for Neutron Activation Measurement Facility
at ARTI, Jeongeup
Jounghwa Lee et al., Korea Atomic Energy Research Institute, Korea

P1-09

Monte Carlo optimization design of simulated workplace neutron fields for typical PWR
workplaces
Faguo Chen et al., China Institute for Radiation Protection, China

P1-10

Improvement of integrated functions of the dose rate model for incident-freetransportation of radioactive materials
Kiwon Jang et al., Department of Radiation Protection & Radioactive Waste Safety, Korea
Institute of Nuclear Safety, Korea

P1-11

Effects of Dose to Remaining Buildings from the Nuclear Power Plant Decommissioning
by Shielding Materials and Thickness
Hyung Woo Seo et al., Decommissioning Technology Group/KHNP CRI, Korea

P1-12

Research and Verification of Radiation Shielding for a Non-destructive Testing Device ofa Coal Mine Belt
Jianyu Fu et al., China Research Institute of Radiation Protection, China

Session P2
P2-01

Analysis of usefulness of obesity risk factors using ROC curve for male workers
Baek Yun Mi et al., Korea Hydro & Nuclear Power Co., Ltd, Korea

P2-02

Development of n-γ mixed fields based on a D-D neutron source for validation of materialfiltered RPLGD for BNCT
Zixu Xu et al., Graduate School of Engineering, Osaka University, Japan

P2-03

Dosimetric comparison of different techniques of intraoperative radiotherapy for spinal
metastasis based on MC simulation
Junyi Liu et al., School of Nuclear Science and Technology, University of Science andTechnology of China, China

P2-04

Development of Computed Tomography-dedicated Mesh-Type Pediatric Phantom
Library: Arms-removed and Arms-raised Posture
Suhyeon Kim et al., Hanyang University, Korea

P2-05

Image

Reconstruction

with

Limited-view-angle

Projection

Data

for

BNCT-

SPECT―Establishment of the Mock-up System and Validation of the Response
Function―
Fan Lu et al., Department of Engineering, Osaka University, Japan
P2-06

Establishing Local Exposure Scenarios for Local Skin Dose Assessment System
Seung Beom Goh et al., Department of Nuclear Enginnering, Hanyang University, Korea

17

P2-07

Numerical simulation for calculating counting efficiencies considering the biodistributionof I-131 in hand-held radiation instrument
Min Seok Park et al., Department of Emergency Medical Research, Korea Institute ofRadiological and Medical Sciences, Korea

P2-08

Mesh-Type Phantom Posture Deformer: A Real-Time GUI for Personalized Phantom
Generation
Haegin Han et al., Department of Nuclear Engineering, Hanyang University, Korea

P2-09

Deep Learning for stable Chromosome Analysis in Bio-Dosimetry
In Kyung Lee et al., RHI, Korea Hydro & Nuclear Power Co., LTD, Korea

P2-10

Deep Learning for UnStable Chromosome Analysis in Bio-Dosimetry
In Kyung Lee et al., RHI, Korea Hydro & Nuclear Power Co., LTD, Korea

P2-11

Enhancing Radiologist Pose Tracking in Interventional Radiology: A YOLOv7 and MultiCamera Approach for Virtual Dose Monitoring System
Sungho Moon et al., Hanyang University, Korea

P2-12

Development of GUI-based Monte Carlo Simulation Code for External Exposure(McSEE): Implementation of Advanced Functions
Hyeonil Kim et al., Department of Nuclear Engineering, Hanyang University, Korea

P2-13

Evaluating Fetal Radiation Exposure: Development and Impact of Pregnant-female Meshtype Reference Computational Phantoms for Accurate Dosimetry
Bangho Shin et al., Department of Nuclear Engineering, Hanyang University, Korea

P2-14

Development of a prototype TL/OSL reader for on-site application in a large-scale
radiological accident
Hyoungtaek Kim et al., Radiation Safety Management Division, Korea Atomic Energy ResearchInstitute, Korea

P2-15

Evaluation of shielding effect of personal dosimeter according to working postures
Min Chae Kim et al., Korea Atomic Energy Research Institute (KAERI), Korea

P2-16

Dose Evaluation of High-energy Photons Leaking from Kindai Reactor
Shintaro Kimura et al., Grad. Sch. Kindai Univ, Japan

P2-17

Comparing the Performance of an Automatic System with Manual Scoring in the
Dicentric Chromosome Assay
Soo Kyung Jeong et al., Dongnam Institute of Radiological & Medical Sciences, Korea

P2-18

Neutron Dose Inside a 50 MeV Cyclotron Vault
Ahn Yunjin et al., Radiation Safety section,Korea Institute of Radiological and Medical
Sciences, Korea

P2-19

Uncertainty Evaluation of calibration of a Beta-ray Irradiation System using the MonteCarlo Method

18

Moonhyung Cho et al., Central Research Institute, Korea Hydro & Nuclear Power Co., Ltd.,
Korea
P2-20

Improvement of the Performance of a Whole Body Counter Using an Artificial Neural
Network
Moonhyung Cho et al., Central Research Institute, Korea Hydro & Nuclear Power Co., Ltd.,
Korea

P2-21

Research on mRNA as a Biomarker of Mixed Neutron and γ -ray Irradiation
Yuan Yayi et al., Department of Radiation Medicine and Environmental Medicine, China
Institute for Radiation Protection, China

P2-22

Establishment and application of the biokinetic model of plutonium liver cells
Shuang Shuang Cui et al., Institute of Radiology and Environmental Medicine, China Institutefor Radiation Protection, China

Session P3
P3-01

Development of a Methodology to Convert the Energy Spectra to Spectra for Channel
Ju Hyung Kim et al., Department of Nuclear Engineering, Hanyang University, Korea

P3-02

Correction of changes in Photopeak Position and FWHM according to Temperature
Changes of CsI(Tl) Detector
Chan Joong Kim et al., Department of Nuclear Engineering of Hanyang University, Korea

P3-03

Evaluation of Output Accuracy and Precision of 3D-Printed Scintillator (RMPS470)
Junehyung Bernask et al., Department of Nuclear Engineering, Hanyang University, Korea

P3-04

Improvement and evaluation of anti-tumor effect by neutron capture therapy using a newelement — rhodium based on beta ray
Xuan Hou et al., Department of Bioengineering, School of Engineering, The University ofTokyo, Japan

P3-05

Evaluation of Properties of Plastic Scintillators Fabricated by 3D Printing
Han Cheol Yang et al., Department of Nuclear Engineering, Hanyang University, Korea

P3-06

Accelerated Monte Carlo-Based Gamma-Ray Spectrometer Modelling for Efficient
Dataset Creation in Radiation Detection and Sensor Technology
Youngsu Kim et al., Korea Atomic Energy Research Institute, Korea

P3-07

Surface contamination meters and monitors based on plastic scintillator–WLS–SiPM
technology produced on SPC “Doza”
Pavel Buzhan et al., Scientific Production Company "Doza", Russian Federation

P3-08

Implementation of Digital Pulse Processing for Neutron Measurement Using Recursive
Algorithms
Jae Chang Kim et al., Department of Nuclear Engineering, School of Hanyang University,

19

KoreaP3-09

Coded Aperture Gamma Camera Prototype Based on CeBr3 Single Crystal PositionSensitive and Spectrometric Detector
Yury Teverovskiy et al., SPC Doza, Russian Federation

P3-10

A study on methods to minimize pulse pile-up and dead time errors in high radiation fields
Sujung Min et al., Department of decommissioning technology development, Korea Atomic
Energy Research Institute, Korea

P3-11

Evaluation of radiation detection properties of liquid scintillators based on modified
scintillation material
Sujung Min et al., Department of decommissioning technology development, Korea Atomic
Energy Research Institute, Korea

P3-12

Development of miniature gamma-ray imager using 3D printed tungsten coded aperture
Joonhyuk Lee et al., Jeju National University, Korea

P3-13

Enhancing Proton Therapy Range Verification: The Multi-Slit Prompt-Gamma Camera
for Improved Conformal Dose Delivery
Youngmo Ku et al., Depaprtment of Nuclear Engineering, Hanyang University, Korea

P3-14

Accelerating Cargo Screening: Development and Evaluation of a Hybrid Gamma Imaging
System for Rapid Detection of Nuclear and Radioactive Materials
Junyoung Lee et al., Department of nuclear engineering, Hanyang University, Korea

P3-15

Addressing Temperature Variations in Outdoor Gamma Imaging: A Novel RegressionBased PMT Gain Optimization Technique
Goeun Lee et al., Department of nuclear engineering, Hanyang University, Korea

P3-16

Development of CZT-based Portable Gamma Imaging System for Korea Customs Service
Jae Hyeon Kim et al., Advanced Radiation Technology Institute, Korea Atomic Energy ResearchInstitute, Korea

P3-17

Feasibility Study on Pulse Shape Discrimination of Fast/Thermal Neutrons, and Gamma
ray with a Single Detector
Chaehun Lee et al., Korea Atomic Energy Research Institute, Korea

P3-18

Application of Optical Fiber Beam Loss Monitor in SSRF
Dong Zhang et al., Department of Radiation Protection and Technical Safety, Shanghai
Synchrotron Radiation Facility, Shanghai Advanced Research Institute, China

P3-19

Time-of-Flight Neutron Bragg Edge Imaging Using a Compact Accelerator Neutron
SourceMahdi Bakhtiari et al., Division of Advanced Nuclear Engineering, Pohang University of
Science and Technology, Korea

P3-20

A machine learning-based peak area calculation algorithm in low background20

measurement
Peng Zhang, Key Laboratory of Particle and Radiation Imaging (Ministry of Education) and
Department of Engineering Physics, Tsinghua University, China
P3-21

Experimental verification of real-time gamma-ray energy spectrum and dose monitor up
to 3 MeV
Nikolaos Voulgaris et al., Division of Sustainable Energy and Environmental Engineering,
Graduate School of Engineering, Osaka University, Japan

P3-22

Remote Radiation Detection Imaging System Mounted on a Drone
Youngjoo Park et al., Department of Biomedical Engineering, School of Korea University,
Korea

P3-23

Development of a Coded Aperture Collimator for Drone-Based Radiation Imaging System
Inbum Lee et al., Department of Biomedical Engineering, Korea University, Korea

P3-24

Simulation of boron-coated honeycomb tube neutron detector with high detection
efficiency
Jianjin Zhou et al., China Institute for Radiation Protection, China

P3-25

Characteristic Determination of 1,800 cc Ion Chamber and Radiation Measurement for
Human Body Scanner
Yong Ho Jin et al., Kyung Hee University, Korea

P3-26

Neutron multiplication and prompt neutron angular distributions of diverse nuclear
material samples
Seonkwang Yoon et al., Advanced Fuel Cycle Technology Development Division, KAERI, Korea

P3-27

Development of a New Low-Energy Neutron Spectrometer Using a Position-Sensitive
Proportional Counter -Examination of Response Functions Including ShieldYu Fujiwara et al., Osaka University, Japan

P3-28

The performance analysis of the laundry monitor
Jing Zhang et al., China Institute for Radiation Protection, China

P3-29

Investigation of a Statistical Spatial Analysis Methodology for Measurement-based Target
Point’s Spatial Dose Rate Estimation.
Ji Woo Kim et al., Department of Nuclear Engineering, Kyung Hee University, Korea

P3-30

Development of Online Directional Dose Equivalent Rate Monitoring System
Yingjing Wei et al., Institute of Nuclear and New Energy Technology, Tsinghua University,
China

P3-31

Design and simulation verification of a coded aperture gamma camera with different
vertical and horizontal fields of view
Yan Li et al., Shanxi Provincial Key Laboratory for Radiation Safety and Protection, CNNC
Key Laboratory for Radiation Protection Technology, China Institute of radiation protection,

21

ChinaP3-32

Application of new UAV radiation aerial survey technology in nuclear emergency
Mei Feng, Department of Health Physics, China Institute for Radiation Protection, China

P3-33

Monte Carlo Simulation for An Underground Radioactivity Measurement
Chen Jikai, Department of Engineering Physics, Tsinghua University, China

P3-34

3D radiation mapping: 3D scene and gamma radiation dose rate measurement data fusion
Hui Li et al., China Institute for Radiation Protection, China

P3-35

Research status and development of neutron detector
Hengchun Zhao et al., China Institute for Radiation, China

Session P4
P4-01

Natural radioactivity in bottled water and associated age-dependent dose and lifetime risk
assessment in Korea
Ji Eun Lee et al., Dongnam Institute of Radiological & Medical Sciences, Korea

P4-02

Optimization of the detection system for real-time monitoring of marine radioactivity
Mee Jang et al., Environmental Radioactivity Assessment Team, Korea Atomic Energy ResearchInstitute, Korea

P4-03

A study on the method of measuring environmental level 137Cs radioactivity in seawaterDong Hye Heo et al., Ionizing Radiation Metrolgy Group/Division of Chemical and Biological
Metrology/Korea Research Institute of Standards and Science (KRISS), Korea|

P4-04

One-Class SVM Based Anomaly Detection for LSTM-Based Radiation Prediction
Sihyun Lee et al., Department of Electronic Engineering

P4-05

An Approach to Calculation of Individual Doses on Explosion Accident Scenarios during
Decommissioning of Nuclear Power Plants
Kwangho Jo et al., Decommissioning Technology Group, KHNP CRI, Korea

P4-06

LSC method for determination of Total-Uranium and Uranium Isotopes in Liquid
Effluents from nuclear fuel cycle facilities
Yang Youkun et al., China Institute for Radiation Protection, China

P4-07

Main Performance Test of Gaseous Effluent Sampling System
Li Qiao et al., China Institute for Radiation Protection, China

P4-08

Gamma nuclides discrimination technology based on fully connected neural network
Xin Liu et al., China Institute for Radiation Protection, China

P4-09

Research on computational efficiency optimization algorithm based on source weight
discrete
Wei Jiafu et al., China Institute for Radiation Protection, China

P4-10

Distribution of Naturally Occurring Radioactive Materials in Groundwater of Jeju Island

22

Sohyeon Lim et al., Department of Chemistry, Jeju National University, Korea
P4-11

Particle Size Distribution of Radioactive Aerosols in Workplace
Feng Zhang et al., China Institute for Radiation Protection, China

Wednesday, July 5
08:30-08:30

Registration

Room 612
08:30-10:00

Invited Talk

Moderator:

Kyo Youn Kim, Hiroyuki Takahashi

4.

Development of a Series of In Vivo Monitoring Systems for Internal Contamination—from
Gamma Spectrum to Dose Assessment
Liye Liu et al., China Institute for Radiation Protection, Shimizu Corporation, China

5.

General Features of PHITS and its Applications to Cosmic-Ray Dosimetry
Tatsuhiko Sato, Japan Atomic Energy Agency, Japan

6.

Development and Applications of Multi-scale Dosimetry with Chinese Reference Phantom
Library
Rui Qiu et al., Department of Engineering Physics, Tsinghua University, China

10:00-10:30

Coffee BreakRoom 612
10:30-12:00

Session O3

Chair:In Jung Kim, Runcheng Liang, Shunsuke Kurosawa

O3-11

Ultra-dense Ceramic Scintillators for Positron Emission Tomography
Sun Il Kwon et al., Department of Biomedical Engineering, University of California Davis, USA

O3-12

Development of Glass Gas Electron Multiplier based Beta Ray Imaging System
Moh Hamdan et al., Department of Nuclear Engineering and Management, Graduate School ofEngineering, The University of Tokyo, Japan

O3-13

Dosimetric Study of a Novel Perovskite Scintillator for High-Energy Radiotherapy
Ning Gao et al., School of Nuclear Science and Technology, University of Science andTechnology of China, China

O3-14

Background Reduction in PET using Quantum Entanglement Property of Annihilation

23

Gamma Rays
Donghwan Kim et al., Department of Nuclear Engineering and Management, Graduate Schoolof Engineering, The University of Tokyo, Japan
O3-15

Radionuclide analysis Using PHITS and Neutron Irradiation in HANARO Concrete
SampleDong Hyuk Shin et al., Department of Nuclear Engineering, Hanyang University, Korea

O3-16

Development of Prototype Backscatter X-ray Imaging System for Security Screening
Geunyoung An et al., Jeonbuk National University, Korea

12:00-13:30

Lunch

13:30-15:00

Session O3

Chair:

Sang Han Lee, Seo Kon Kang, Runcheng Liang

O3-17

Development of Background Simulation Technique for Radiation Portal Monitors at
Airports and Ports in South Korea
Hayoung Sim et al., Department of Applied Plasma and Quantum Beam Engineering, Jeonbuk
National University, Korea

O3-18

Performance Estimation of PVT Scintillator-based Seafood Monitoring System for
Screening of Radioactive Materials
Hyosang Choi et al., Jeonbuk National University, Korea

O3-19

Radiation Hardness Assurance Testing of a 40-nm Microprocessor Applied in Nuclear
Industry
Runcheng Liang et al., China Institute for Radiation Protection, China

O3-20

Scintillation Properties of Cr-doped α-Al2O3 Crystals for the Decommissioning
Monitoring System
Shunsuke Kurosawa et al., Tohoku University & Osaka University, Japan

O3-21

Research on the performance of SiC thermal neutron detector based on metal-oxidesemiconductor structure
Lei Ren et al., Hefei Institute of Physical Science, Chinese Academy of Sciences, China

O3-22

Proposal of Method for Estimation of Gamma Source Direction based on Self-shielding
Profile in a Multipixel Array Detector
Atsushi Mukai et al., Department of Energy Engineering, Graduate School of Engineering,
Nagoya University, Japan

15:00-15:30

Coffee Break24

15:30-17:00

Session O3

Chair:Ki Taek Han, Haijing Jin, Tetsuo Iguchi

O3-23

Fuel debris mapping tool using an Unmanned Underwater Vehicle to grasp fuel debris
submerged in Fukushima Daiichi Nuclear Power Plant
So Kamada et al., National Maritime Research Institute, Japan

O3-24

Fabrication of Transition Edge Sensors with faster response time
Takeshi Jodoi et al., Department of Engineering, The University of Tokyo, Japan

O3-25

Muography-based method of subsidence risk assessment for underground tunnel
construction
Naoya Okamoto et al., IGSGS, Kyushu University, Japan

O3-26

Waveform discrimination by fitting derivative of synchronized ideal normalized curves,dSINC fit
Martin Andersson et al., Visible Information Center, Inc., Japan

O3-27

Development of Dual Energy X-ray Security Scanner for Improved Organic Material
Detection: A Monte Carlo Study
Junsung Park et al., Department of Applied Plasma and Quantum Beam Engineering, Jeonbuk
National University, Korea

O3-28

Characteristic evaluation of gamma-ray spectrometer using an unfolding method withrefined response functions
Masateru Hayashi et al., Advanced Technology R&D Center, Mitsubishi Electric Corporation,Japan

18:00-21:00

Conference Dinner

Room 606
10:30-12:00

Session O1

Chair:

Chang Ho Shin, Hengguan Yi, Tetsuo Iguchi

O1-03

A Preliminary Study on a GPU Accelerated Monte Carlo Dose Calculation Method for
BNCT
Yuxin Wang et al., School of Nuclear Science and Technology, University of Science andTechnology of China, China

O1-04

Radiation Exposure Assessment for Outdoor Large-Structure Tests in Industrial
Radiography
Junik Cho et al., Department of Nuclear Engineering, Hanyang University, Korea

O1-05

Study on induced radioactivity of different materials in the proton therapy facility

25

Yue Yu et al., School of Nuclear Science and Technology, University of Science and Technology
of China, China
O1-06

Three-dimensional dynamic γ radiation field calculation method for digital simulation ofradiation protection
Haijing Jin et al., China Institute for Radiation Protection, China

O1-07

The Research on Neutron Shielding Performance and The Secondary γ Dose Simulationof Tungsten Boride Material
Yi Han et al., China Institute for radiation protection, China

12:00-13:30

Lunch

13:30-15:00

Session O1 & O4

Chair:

Hyoungtaek Kim, Jing Kang, Hiroyuki Takahashi

O1-08Monte Carlo Analysis Considering Temperature Change of Metal Alloy Materials in the
In-Core Irradiation Test
Hae Sun Jeong et al., HANARO Utilization Division, KAERI, Korea

O1-09

High frequency pulse emission characteristics and application of carbon nanotube cold
cathode X-ray source
Sheng Lai et al., Department of Nuclear Science and Technology, Nanjing University of
Aeronautics and Astronautics, China

O4-04

Numerical analysis and discussion on the implementation of airborne radioactive emission
standards of nuclear power plant radiation protection standard (GB6249-2011)
Fu Shen, China Institute for Radiation Protection, China

O4-05

The content level and particle size distribution of 210Po in Taiyuan environmental aerosol
Pengxiang Li et al., China Institute for Radiation Protection, China

O4-06

Study on the Influence of Large Natural Ventilation Cooling Towers on AtmosphericDiffusion at Offshore NPPs
Junfang Zhang, Department of Nuclear Environmental Science, China Institute for Radiation
Protection, China

15:00-15:30

Coffee Break

15:30-17:00

Session O4

Chair:

Mee Jang, Jing Kang

O4-07

Concentration behavior and radiation dose of 137Cs in zebrafish
Shaofei Cao et al., Institute of Nuclear Environmental Science, China Institute for Radiation

26

Protection, China
O4-08

Preliminary Study on Separation and Concentration for Large Volume of Krypton by
Improved Gas ChromatographXuyuan Ma et al., China Institute for Radiation Protection, China

O4-09

Analysis of Interference between Tritium and Carbon -14 Measurement
Li Bao et al., Key Laboratory of Radiation Environment & Health of the Ministry of Ecology
and Environment, China Institute for radiation protection, China

O4-10

Study on Optimal Allocation Method of Dose Constraint Value in Nuclear Facilities
Hailong Chen et al., Key Laboratory of Radiation Environment & Health of the Ministry of
Ecology and Environment, China Institute for Radiation Protection, China

O4-11

The Graded Approach in the Radiation Environment Management of NORM in China
Wang Yan et al., China Institute for Radiation Protection, China

18:00-21:00

Conference Dinner

Room 604
10:30-12:00

Session O8

Chair:

Haeryong Jung, Chao Gao, Takashi Nakamura

O8-03

Research on the key problems of the new smelting and purification process of metal
radioactive pollutants
Kewei Jiang, China Institute for Radiation Protection, China

O8-04

The Study of Mechanical wear principle and dynamics of graphite spheres in HTR-PMhigh temperature gas cooled reactor
Lixiao Guo, China Institute for Radiation Protection, China

O8-05

Development of Microwave Drying Technology for Radioactive Wet Waste Treatment inDrum
Meng Zihe et al., The Department of Radioactive Waste Management, China Institue for
Radiation Protection, China

O8-06

Development and verification of portable incineration technology for low-level radioactivewaste
Chu Haoran et al., China Institute for Radiation Protection, China

O8-07

Research on Si-Al base Geopolymers Formulation of Simulated Radioactive Fluoride
Liquid Wastes
Li Zhang et al., China Institute for Radiation Protection, China

O8-08

The Study on Microwave Drum Drying for Radioactive Wet Waste

27

Gao Chao et al., Radioactive Waste Management Department, China Institute for Radiation
Protection, China

12:00-13:30

Lunch

13:30-15:00

Session O8

Chair:

Jeongken Lee, Chao Gao

O8-09

A synthesized bifunctional strongly basic anion exchange resin for ReO4-/TcO4- removal
Yujia Kang et al., School of Nuclear Science and Technology, University of Science andTechnology of China, China

O8-10

Research and application of on-site calibration technology for a radioactive waste drum
NDA inspection system
Hengguan Yi et al., Shanxi Provincial Key Laboratory for Radiation Safety and Protection,
China & Key Laboratory of Radiation Protection of CNNC, China & China Institute for
Radiation Protection, China

O8-11

Activation status of metallic components in the PET-cyclotron revealed by destructive andnon-destructive analysis
Go Yoshida et al., High Energy Accelerator Research Organization (KEK), Japan

O8-12

Efficient Adsorption of Sr2+ from High Level Liquid Waste by 4A Zeolite
Xiang Qin et al., China Institute for Radiation Protection, China

O8-13

Dose estimation for radioactive metal waste handling workers using dose conversion
factors of ICRP Publications 119, 141, and EPA FGR No. 11
Jin Ho Son et al., Department of Nuclear Engineering, Chosun University, Korea

O8-14

Speciation simulation and sorption mechanism of 238Pu in waste repository groundwater
Chen Chao et al., Department of Nuclear Environmental Science, China Institute for Radiation
Protection, China

15:00-15:30

Coffee Break

15:30-17:00

Session O8 & O5

Chair:

Manhee Jeong, Chao Gao

O8-15

Gaseous radioactive iodine capture of microporous molecular sieves supported silver:
BEA, MFI and CHA
Zexiang Chen et al., Key Laboratory of Nuclear Air Purification Technology, China Institute for
Radiation Protection, China

O8-16

The improvement in the iodine capture property of silica-based aerogel by introducing

28

three-dimensionally ordered microporous
Sen Chang et al., Department of Environment Engineering and Technology, China Institute for
Radiation Protection, China
O8-17

Thermodynamic Analysis of Effect of Waste Resin Irradiation Gas on Mineral
Composition of Ordinary Portland Cement
Han Xu et al., The Department of Radioactive Waste Management, China Institute for Radiation
Protection, China

O5-02

Radiological Risk Assessment of Safety Exit Sign Containing Tritium
Yeijin Bang et al., Department of Applied Plasma and Quantum Beam Engineering, Jeonbuk
National University, Korea

O5-03

Radiation Dose Assessment of Thorium-containing Gas Mantle for Predictable Exposure
Scenarios
Jimin Shin et al., Department of Applied Plasma and Quantum Beam Engineering, Jeonbuk
National University, Korea

18:00-21:00

Conference Dinner

Room 608
Poster Session
08:30-12:00
12:00-13:30
13:30-15:00
15:00-17:00
18:00-21:00

Poster Display
Lunch
Poster Presentation
Poster Display
Conference Dinner

Session P4
P4-12

Pretreatment Method and Ash Fresh Ratio Calculation of Biological Sample Analysis in
Radiation Environment Monitoring
Jinhao Li et al., China Institute for Radiation Protection, China

P4-13

Cooperation on Radiation Measurements for Cross-Border Accidents (2) Application ofmachine learning conversion method using monitoring data from Fukushima as teachingdata
Miyuki Sasaki et al., Japan Atomic Energy Agency, Japan

P4-14

Cooperation on Radiation Measurements for Cross-Border Accidents (1) A CZT detectorbased airborne survey analysis in high dose rate environment

29

Sungyeop Joung et al., Environmental Safety Technology Research Division, Korea Atomic
Energy Research Institute, Korea
P4-15

Feasibility Study of Directly Measuring Radon Concentration in Air Using TPC
Qiuli Zhang, Department of Engineering Physics, Tsinghua University, China

P4-16

Life cycle greenhouse gas emissions from nuclear power chain in China
Feifei Wu et al., China Institute for Radiation Protection, Key Laboratory of Radiation
Environment & Health of the Ministry of Ecology and Environment, China

P4-17

Study on Optimal Control of Liquid Effluent Emission from Small Nuclear Power Reactor
Meng Wang et al., China Institute for Radiation Protection, Key Laboratory of Radiation
Environment & Health of the Ministry of Ecology and Environment, China

P4-18

Determination of 226Ra in drinking water in Southern China by Liquid Scintillation
Counting
Chen Guo et al., China Institute for Radiation Protection, China

P4-19

Combined transuranic-nuclide analysis.of aerosol samplese
Ke Du et al., China Institute for Radiation Protection, China

P4-20

The Content of Tritium and Carbon-14 in Environmental Medium around Shidaowan
High Temperature Gas Cooled Reactor NPPQin Nan Song, China Institute for Radiation Protection, China

P4-21

Assessment of Uncertainty in Collecting Radioactive Soil Samples by Sampling andAnalytical Quality Control（（SAX））
Ruijun Wang et al., Key Laboratory of Radiation Environment & Health of the Ministry of
Ecology and Environment, China Institute for radiation protection, China

P4-22

Evaporation behavior of polonium-210 from lead-bismuth eutectic at low temperature
Huiyun Xue et al., University of Science and Technology of China, Anhui Province, China

P4-23

Wind tunnel study of effects of thermal stratification on turbulent flow and diffusion overa three-dimensional hill
Yun-peng Li et al., China Institute for Radiation Protection, China

P4-24

Background Spectrum Simulation Development and Verification for In-Situ Measurement
Technology Application
Byungchae Lee et al., R&D Institute, Sae-An Enertech Corp, Korea

P4-25

Sensitivity Analysis of Distribution Coefficient by Nuclide According to Landfill of NORM
Waste
Seong Yeon Lee et al., Department of Nuclear Engineering, Kyung Hee University, Korea

P4-26

A Study on Radiological Characterization for Alpha Source Terms in Decommissioning
Nuclear Power Plants
Seong Jun Kim et al., Department of Nuclear Engineering, School of Chosun University, Korea

30

P4-27

Deep learning-based spectrum-dose conversion model for KAERI MARK-M detection
systemEunjoong Lee et al., Environmental Safety Technology Research Division, Korea Atomic Energy
Research Institute, Korea

P4-28

Study of the radiation effects of radioactive residues in soil at a decommissioning site
Zhi Xiang Yu, China Institute for Radiation Protection, China

P4-29

Analysis of Carbon-14 Radioactive Effluents Released from Korean Nuclear Power Plants
and Its Resulting Public DoseHwa Pyoung Kim et al., Chosun university Department of Nuclear Engineering, Korea

P4-30

Specification for monitoring of uranium aerosol concentration in the air at workplaces ofuranium processing and fuel fabrication facilities
Baoli Wu et al., China Institute for Radiation Protection, China

P4-31

Effects of Scenarios and Source Types on Doses for the Reuse Buildings from theDecommissioning of Nuclear Facilities
Hyung-Woo Seo et al., Decommissioning Technology Group, KHNP CRI, Korea

P4-32

Review of Dose Estimation Considering Decay Correction for the Decommissioning of
Nuclear Power Plants
Hyung-Woo Seo et al., Decommissioning Technology Group, KHNP CRI, Korea

P4-33

Related research on ecological environmental impact assessment around nuclear fuelreprocessing plant
Gong Wenjing et al., China Institute for Radiation Protection, China

Session P5
P5-01

Impact of vendor-agnostic deep learning-based CT denoising technique on patient’s
radiation dose reduction in a multi-vendor CT clinical environment
Ahyeong Lee et al., a Department of Applied Bioengineering, Graduate School of Convergence
Science and Technology, Seoul National University, Korea

P5-02

Review on Radiation Protection Program during Decommissioning of NPP
Hye Jin Jung et al., Decommissioning Technology Group, KHNP CRI, Korea

P5-03

Current Status Analysis to Optimize Shielding Design Criteria for Radiotherapy Facilities
using Linear Accelerators
Sang Hyun Park et al., Korea Institute of Nuclear Safety, Korea

P5-04

Review of the radiation dose of aircraft crew due to cosmic radiation
Byungmin Lee et al., Korea Institute of Radiological and Medical Sciences, Korea

P5-05

A Review on Factors Necessary for Estimating Worker Exposure Dose for
Decommissioning Nuclear Power Plants

31

Gi-Lim Kim et al., Decommissioning Technology Group, KHNP CRI, Korea
P5-06

Comparison of Dose Assessment of Off-link Public According to Radioactive Waste
Transportation by Computer Code
Minwoo Kwak et al., Department of nuclear engineering, Kyunghee University, Korea

P5-07

Assessment of Accident Probability for Overland Transportation of Spent Nuclear Fuel in
Korea
Jooyub Lee et al., Department of Nuclear Engineering, Seoul National University, Korea

P5-08

Evaluation of Preliminary Risks Derived for the Decommissioning of Nuclear Facilities
Hyung-Woo Seo et al., Decommissioning Technology Group, KHNP CRI, Korea

Session P6
P6-01

Development of Quantitative Assessment System for Evaluation of Training Competencein Virtual Reality-based LINAC QA Training
Jongeun Kim et al., Department of Nuclear Engineering, Hanyang University, Korea & Proton
Therapy Center, National Cancer Center, Korea

P6-02

Comparison of Radiation Regulatory Systems in Selected Countries : Focusing on
Differences and Adaptations
Euna Lee et al., Department of Nuclear Engineering, Hanyang University, Korea

P6-03

Analysis of Radiation Monitoring System for Imported Cargo in Korea, the United States,
the United Kingdom, and Japan
Choi Wo Suk et al., Department of Chosun University, Korea

P6-04

Student Training Program for Radiation Safety in Hanyang University: HY-Radiedu
program and Radiation Safety technology conference
Euna Lee et al., Department of Nuclear Engineering, Hanyang University, Korea

Session P7
P7-01

Review of Establishing EPZ through GIS tools
Bongseok Kim, Nuclear Emergency Preparedness Team, Korea Atomic Energy ResearchInstitute, Korea

P7-02

Development of GUI Program for Dose Assessment for Skin Contamination in Radiation
Accident
Lee Seung Je et al., National Radiation Emergency Medical Center, Korea Institute ofRadiological and Medical Sciences (KIRAMS), Korea

P7-03

Derivation of Considerations for Development of Victim Management Procedure in
Radiation Emergency
Kihoon Kim et al., Korea Institute of Radiological & Medical Sciences, Korea

32

P7-04Development of a Screening Method for Contamination in Urine Samples during
Radiation Emergencies
Seokwon Yoon et al., Korea Institute of Radiological and Medical Sciences, Korea

P7-05

Radiation exposure dose reduction in radiation emergency medicine
Jaeryong Yoo et al., Korea Institute of Radiological and Medical Sciences, Korea

P7-06

Analysis of Mountain-Valley Breeze Morphology Affecting the Diffusion and Distributionof Pollutants at an Inland Nuclear Facility
Huang Sha et al., China Institute for Radiation Protection, China

P7-07

Comparison of diffusion simulation results of JRODOS and NACADOS at a coastal plantsite in China
Zhao Duoxin et al., China Institute for Radiation Protection, China

P7-08

Emergency Action Levels Development Study for Research Reactors
Ilje Cho et al., Nuclear Emergency Preparedness Section, Korea Atomic Energy ResearchInstitute, Korea

P7-09

Rapid determination of Pu, Am and Cm in urine sample by tandem quadrupole ICP-MS
Yang Wu et al., China Institute for Radiation Protection, China

P7-10

Effects of Neutron, γ-Ray, and Neutron–γ-Ray Mixed Radiations on Chromosomal
Aberrations in Peripheral-Blood Lymphocytes
Zhang Zhongxin et al., Radiological and Environmental Medicine Department of China
Institute for Radiation Protection, China

P7-11

Single-column Chromatography Separation for Simultaneous Determination of Pu, Np, U,
Th in Small-volume Urine for Emergency Radioassays
Maoyi Luo et al., China Institute for Radiation Protection, China

P7-12

Establishment of the Dose-response Calibration Curve for Semi-automatic Analysis ofDicentric Chromosome
Qianqian Meng et al., China Institute for Radiation Protection, China

P7-13

Role of TLR4/NF-κB signaling pathway in hydrogen-rich water improving acuteradiation-induced brain injury in rats
Mengya Liu et al., Department of Radiology and Environment Medicine, China Institute for
Radiation Protection, China

P7-14

Evaluation of the Portable Radiation Portal Monitor’s Usability for Contamination
Screening of Large Groups in a Radiological Emergency
Young Min Lee et al., Dept. of Nuclear Emergency Preparedness, Korea Institute of Nuclear
Safety, Korea

Session P8

33

P8-01Development optimized evaluation technology for work safety linking location tracking
and in situ measurement in decommissioning facilities
Sang Bum Hong, Decommissioning technology development division, Korea Atomic Energy
Research Institute, Korea

P8-02

Component analysis for solidification after incineration alternative thermal treatment
process of dry active wastes
Ji Hyun Yu et al., Sae-An Enertech Corporation, Korea

P8-03

Measurement of Thermal Neutron Fluence Rates of a Proton Therapy Facility Using a
Superconducting Synchrocyclotron - in the Case of IBA ProteusOne at Sapporo Kojinkai
Memorial Hospital, Japan
Hajime Nakamura et al., Radiation Science Center, High Energy Accelerator Research
Organization, Japan

P8-04

Accident Scenario Analysis in Radioactive Waste Treatment Facility
San Chae et al., Korea Atomic Energy Research Institute (KAERI), Korea

P8-05

Development of Radwaste Management System based on ICT technology
Jongjin Kim et al., Korea Atomic Energy Research Institute (KAERI), Korea

P8-06

A Study on the Packaging Measures for Radioactive Waste of NPP Decommissioning
Jun Ki Baik et al., Central Research Institute, Korea Hydro & Nuclear Power (KHNP), Korea

P8-07

Study on the Treatment of Radioactive Spent Resin by Wet Oxidation
Yahui Xi, China Institute for Radiation Protection, China

P8-08

Radioactivation Investigation in Synchrotron-Type Proton Therapy Facilities
Hiroshi Matsumura et al., Radiation Science Center, KEK, Japan

P8-09

Analysis of Low-Level Radioactive Concrete Waste using Gamma Spectroscopy applying
with Efficiency Transfer
Kyunghun Jung et al., Korea Atomic Energy Research Institute, Korea

P8-10

Preliminary study of the effect of dominant Co-60 in gamma spectrum of radioactive waste
Sooyeon Lim et al., Central Research Institute of KHNP, Korea

P8-11

Dose Assessment for Regulatory Clearance of Organic Liquid Radioactive Waste at
KAERIKi Baek Shin et al., Korea Atomic Energy Research Institute, Korea

P8-12

Degumming test of radioactive filter element
Yongtai Yuan et al., China Institute for Radiation Protection, China

P8-13

Recommendations for the Supervision of Mobile Radioactive Waste Incinerator
Su Ziqiang et al., China Institute for Radiation Protection, China

P8-14

In-situ evaluation of neutron flux using activated reaction of 55Mn(n,γ)56Mn,
23Na(n,γ)24Na on the component of a tandem accelerator

34

Go Yoshida et al., High Energy Accelerator Research Organization (KEK), Japan
P8-15

Derivation the activity of 129I limit for near surface disposal
Yangjun Zhao et al., China Institute for Radiation Protection, China

P8-16

Comparison of Cost about Equipment for Spent Nuclear Fuel Management by Dry Storage
Method in Korea
Hyeok Jae Kim et al., Department of Nuclear Engineering, Kyung Hee University, Korea

P8-17

Analysis of Recyclability of Concrete Wastes Generated from Nuclear Facilities during
Decommissioning
Ji Ung Kim et al., Department of Nuclear Engineering, Chosun University, Korea

P8-18

Radiochemical analysis for difficult-to-measure radionuclides in radioactive samples
Myung Ho Lee, KRIST, Korea

P8-19

Establishment of radioactive waste full cycle history management system using QR-code
Jin-Woo Lee et al., KAERI (Korea Atomic Energy Research Institute), Korea

P8-20

A Case Study on the Removal and Disposal of RPV from the La Crosse Boiling Water
Reactor Decommissioning Experience.
Jihwan Yu et al., Korea Hydro & Nuclear Power (KHNP) Central Research Institute, Korea

P8-21

Adsorption Kinetics of Key Nuclides on Geotechnical Media Form Reprocessing Plant
Tian Xie, China Institute for Radiation Protection, China

P8-22

Study on Rock-Soil Dispersion of Vadose Zone and Aquifer in Typical Uranium Mining
Area in Northern Guangdong
Zhao Ming Wang et al., China Institute for Radiation Protection, China

P8-23

Mössbauer Investigation of Low Carbon Steel-Sodium Bentonite Mixture Aged by γ-Rays
Irradiation
Dong Liang, Department of Radioactive Waste Management, China Institute for Radiation
Protection, China

Thursday, July 6
08:00-08:30

Registration

Room 612
08:30-10:00

Invited Talk

Moderator:

Hee Seock Lee, Qun Liu

7.

Multi-scale modeling in Radiation Therapy - Bridging the gaps between Physics andBiology

35

Jan Schuemann, Massachusetts General Hospital & Harvard Medical School, USA
8.

Dose Coefficients of the ICRP for external exposures – Current Development and Status
Nina Petoussi-Henss, Federal Office for Radiation Protection, Germany

9.

Radiological Protection in Nuclear Medicine: ICRP Activities and Guidelines for PET/CT
and Radionuclide Therapy
Keon Wook Kang, Seoul National University, Korea

10:00-10:30

Coffee Break

10:30-11:30

Session O3

Chair:

Jung Ho Kim, Haijing Jin, Hiroyuki Takahashi

O3-29

Multi-Class Radioisotope Identification using Convolutional Neural Networks with TwoDimensional Input Data
Yong Hyun Kim et al., Department of Nuclear Engineering, Hanyang University, Korea

O3-30

Pile-up Separation and Pulse Shape Discrimination for an Organic Scintillation Detector
using Multi-task Learning
Byoungil Jeon et al., Korea Atomic Energy Research Institute, Korea

O3-31

Application of a novel scintillation gel-based detector for 3D dosimetry in radiotherapy
Hua Li et al., China Institute for Radiation Protection, China

O3-32

Building and performance test of X-ray communication system based on grid-controlled
modulated X-ray tube and LYSO-SiPM X-ray receiverJunxu Mu et al., Department of Nuclear Science and Technology, Nanjing University of
Aeronautics and Astronautics, China

11:30-12:00

Closing Session

12:00-13:30

Lunch

Room 606
10:30-11:30

Session O4

Chair:Min Cheol Han, Jing Kang

O4-12

Models and Parameter studies for Retrospective Evaluation of Nuclear Facilities
Jing Kang et al., China Institute for Radiation Protection, Key Laboratory of Radiation
Environment & Health of the Ministry of Ecology and Environment, China

O4-13

Application of Deconvolution Method to Alpha-ray Energy Spectrum to Improve36Performance of Radiation Airborne Particulate Monitor
Makoto Sasano et al., Mitsubishi Electric Corp., Japan
O4-14

CFD Simulation of Resuspension Release of Radioactive Aerosol Particles fromContaminated Sites
Jia Chen Chen et al., China Institute for Radiation Protection, China

O4-15

Parameter sensitivity analysis of different ecological environments for a tritium
environmental transfer model
Junyu Cheng et al., China Institute for Radiation Protection, China

12:00-13:30

Lunch

Room 604
10:30-11:30

Session O5 & O6

Chair:Tae Young Kong, Junfang Zhang, Tetsuo Iguchi

O5-04

Assessment of Source Terms and Potential Doses Due to Steam Generator Tube Ruptureof VVER-1200 at El Dabaa Nuclear Power Plant in Egypt
Waad Saleh et al., Department of radiation safety and environmental protection, Kepco nuclear
graduate school, Korea

O5-05

Thermal Safety Assessment of Transport Containers for Radioactive Materials
Changwu Wang et al., China Institute for Radiation Protection, China

O5-06

Discussion of “Application of knowledge management in radioactive waste management”
Enbang Liu et al., China Institute For Radiation Protection, China

O6-02

Report on the Association for Radiation Science and Applications as a Collaboration of
Academia and Industry in Japan
Tetsuo Iguchi et al., Nagoya University, Japan

12:00-13:30

Lunch

Korea Institute of Radiological & Medical Sciences (KIRAMS)
13:30-17:00

Technical Tour

Friday, July 7

37

Room 606
08:30-12:00

PHITS Tutorial

12:00-13:30
13:30-18:00

Lunch
PHITS Tutorial

38

Invited speakers List
No.

Title

Affiliation

1

Current Activities of the International Commission on Radiological Protection

Authors
Kun-Woo Cho

Korea Institute of Nuclear Safety

2

Review and Revision of the
System of Radiological Protection: RP for the Next Generation

Christopher H.
Clement

International Commission on
Radiological Protection

3

Fast high granular optical Transition Edge Sensor with parallel
biasing scheme for quantum
applications

Hiroyuki Takahashi
and Yuki Mitsuya

The University of Tokyo

4

Development of a Series of In
Vivo Monitoring Systems forInternal Contamination—from
Gamma Spectrum to Dose Assessment

Liye Liu, Yunshi Xiao,
Ri Zhao, Xiaodun Li,
Yichong Liu,
Qinjian Cao,
Yuan Zhao

China Institute for Radiation
Protection

5

General Features of PHITS and
its Applications to Cosmic-Ray
Dosimetry

Tatsuhiko SATO

Japan Atomic Energy Agency

6

Development and Applicationsof Multi-scale Dosimetry with
Chinese Reference Phantom
Library

Rui Qiu1,* ,Zhen Wu2,
Ankang Hu1,
Xiyu Luo1,
Jiahao Wang1,
Ziyi Hu1,
Shuchang Yan1,
Wanyi Zhou1,
Junli Li1

Department of Engineering
Physics, Tsinghua University,
Beijing, China1, Nuctech Company
Limited, Beijing, China2

7Multi-scale modeling in Radiation Therapy - Bridging thegaps between Physics and Biology

Jan Schuemann

Massachusetts General Hospital& Harvard Medical School

8

Dose Coefficients of the ICRP
for external exposures – Current Development and Status

Nina Petoussi-Henss

Federal Office for Radiation
Protection

9

Radiological Protection in Nuclear Medicine: ICRP Activities
and Guidelines for PET/CT andRadionuclide Therapy

Keon Wook KANG

Seoul National University

39

Invited speakers

01

Current Activities of the International Commission on
Radiological Protection
Kun-Woo Cho
ICRP Main Commission member
Korea Institute of Nuclear Safety, Seoul, Korea
The International Commission on Radiological Protection (ICRP) is an independent international organisation that advances for the public benefit the science ofradiological protection, in particular by providing recommendations and guidanceon all aspects of protection against ionising radiation. ICRP is a charity (not-forprofit organisation) registered with the Charity Commission of England and Wales(registration number 1166304). ICRP is a community of more than 350 globallyrecognised experts in radiological protection science, policy, and practice fromabout 50 countries.
ICRP was established in 1928 at the second International Congress of Radiology.
At the time it was called the International X-ray and Radium Protection Committee(IXRPC), but was restructured and given its present name in 1950. ICRP is comprised of the Main Commission, the Scientific Secretariat, four standing Committees (Committee 1 on Effects, Committee 2 on Doses, Committee 3 on Medicine, and Committee 4 on Application), and a series of Task Groups. There are
now 32 active Task Groups that are working to produce the drafts of ICRP publication. When establishing new Task Groups, ICRP calls for expressions of interest for experts to complete the membership. Current opportunities to join ICRPTask Group can be found at https://www.icrp.org/ page.asp?id=584.
ICRP has published more than one hundred and fifty publications on all aspectsof radiological protection. Most address a particular area within radiological protection, but a handful of publications, the so-called fundamental recommendations, each describe the overall system of radiological protection. ICRP Publication 103 is the currently applicable fundamental recommendations. The primary
aim of the Commission’s fundamental recommendations is to contribute to an
appropriate level of protection for people and the environment against the detrimental effects of radiation exposure without unduly limiting the desirable humanactions that may be associated with such exposure.
ICRP has recently established a mentorship programme to engage university students and early-career professionals and scientists as mentees in ICRP Task Groups
with the guidance of an ICRP member as mentor. Mentees may come from educational, governmental, private, or any other organisation. This is a part-time voluntaryarrangement, with mentees continuing to work at their home organisation most of thetime. Current mentorship opportunities can be found at https://www.icrp.org/page.
asp?id=465.

40

02 Invited speakers
Review and Revision of the System of Radiological Protection: RP
for the Next Generation
Christopher H. Clement
International Commission on Radiological Protection

The International Commission on Radiological Protection (ICRP) is undertaking
a review and revision of the System of Radiological Protection that will update
the current general recommendations in ICRP Publication 103 [ICRP Publication
103. Ann. ICRP 37 (2-4), 2007]. This was launched with the open-access paper‘Keeping the ICRP recommendations fit for purpose’ [C Clement et al 2021 J.
Radiol. Prot. 41 1390, DOI 10.1088/1361-6498/ac1611)] which outlined ICRP’s
initial views on areas of the System that may need attention but was fundamentally an invitation for open collaboration.
Follow-up actions have included ICRP’s Future of RP Digital Workshop held 14
October – 3 November 2021, which included 20 live-streamed and 43 on-demand presentations, with approximately 1500 individuals from 100 countries participating, ICRP’s 6th International Symposium on Radiological Protection held in
Vancouver in November 2022, and many other events and opportunities bothonline and in-person.
So far, much of the focus has been on wide and open collaboration to establish
which elements of the System most need review. Based on extensive feedback,
ICRP is confident that all major elements have been identified but will continue
to reassess this throughout the process. The focus is now on the extensive programme of work covered by about 30 ICRP Task Groups each addressing a specific element of the System, and approximately 18 additional topics identified but
not yet initiated. Towards the end of the process, the next General Recommendations will be developed based on the outcome of this work.
ICRP’s mission is to promote RP, primarily by developing recommendations.
However, ICRP is not the owner of the System, or even a user; the System is forthe entire RP community and all those who benefit from it. Therefore, ICRP is
managing the review and revision with a commitment to inclusiveness, accessibility, and transparency, to ensure everyone who wants to contribute can do so,and to benefit from a wide variety of perspectives.
The System of Radiological Protection is robust and has performed well, however, it must adapt to address changes in science and society to remain fit for
purpose. The next General Recommendations, which should be published in theearly 2030’s, will be the basis for radiological protection standards, legislation,and practice for several decades to come.

41

Invited speakers

03

Fast high granular optical Transition Edge Sensor with parallel biasing scheme for quantum applications
Hiroyuki Takahashi and Yuki Mitsuya
The University of Tokyo, Tokyo, Japan

Optical Transition Edge Sensors (TESs) are promising quantum sensors, whichenable single photon counting of infrared photons. It is useful for optical quantum
computing as well as an infrared photon spectroscopy. However, TESs sufferfrom their poor counting rate characteristics compared with superconducting
nanowire devices. Pixelated Transition Edge Sensors are usually operated in an
individual bias and readout scheme. We have proven that the parallel biasing
scheme works for the small pixel TES devices, where pixelated TESs are all connected to the same electrode and operate in parallel. In this biasing scheme, thesystem can be simplified. We have developed a way to identify the pixel throughslight change of signal characteristics for each pixel. Now we are trying to apply
the scheme to speed up the device operation. In this context, the identical signal
feature means identical response of the pixel and we can group signals and use
them in one large detector. Then we can greatly simplify the system and substantially improve the performance with high granularity approach. Apparently, we
saw the improvement of the risetime of our pixelated TES if we scale down the
pixel size from 100um to 20um. We have demonstrated the simultaneous operation of two rows of 20 pixels (total 40 pixels) with this scheme. We are presently
developing an optical TES for quantum computing and the parallel biasing approach can improve the speed, which is important for optical applications. The
small heat capacity leads to the large signal amplitude as well. Actually, it is very
difficult to operate high granular pixelated detector with individual readout
scheme but this scheme can work even for 500nm scale pixels because of thesimplicity. We have fabricated a parallel slit type array of such a small-scale
device (total size 12um x 12um) with the parallel biasing scheme. The device has
been successfully operated and detected pulse signals indicate a better signal
rise time due to the smaller sensitive area of the pixel.

42

04 Invited speakers
Development of a Series of In Vivo Monitoring Systems for Internal
Contamination—from Gamma Spectrum to Dose Assessment
Liye Liu, Yunshi Xiao, Ri Zhao, Xiaodun Li, Yichong Liu, Qinjian Cao, Yuan Zhao
China Institute for Radiation Protection, Taiyuan, China
Airborne radioactive material can cause internal exposure due to intakes from inhalation, ingestion, wound and other routes. That means, whether under normal working
conditions or accident conditions, aerosol radiation protection has always been thefocus and difficult point of radiation protection research. Firstly, this paper briefly
introduces the research overview of the purification, simulation, monitoring, protection and dose assessment of airborne radioactivity by China Institute for Radiation
Protection in recent years. Then, this paper focuses on the research progress of invivo monitoring technology on human internal contamination, as well as some issues
that need to be paid attention to in the future.
In order to further improve the individual monitoring capabilities for internal exposurein China, China Institute for Radiation Protection has carried out a series of targeted
research in recent years. In terms of γ spectra analysis, NASAS, a special γ spectra
analysis software for NaI whole body counters was developed, and its algorithm cansignificantly improve the analysis ability of overlapping peaks, low energy peaks andcontinuum subtraction. In terms of internal dose assessment, INDOSE, an internal
dose assessment software with biokinetic models, committed equivalent dose andeffective dose calculation function, was developed. In terms of in vivo monitoring
equipment, STWBC-1, a standing NaI whole body counter, and a vehicle-mountedinternal contamination monitoring system were developed. These counters were installed with Simple-in vivo, an in vivo monitoring software integrating functions suchas γ spectra acquisition, γ spectra analysis, internal dose estimation, and individual
dose management. At present, an unattended monitoring system has being developed and installed at an internal dose monitoring laboratory of Chinese nuclear power
plant, which allows the measured individual to complete the whole body counting
process autonomously. In terms of calibration technology, Chinese torso simulation
physical phantom CIRP-Torso and the Chinese reference whole body physical phantom CIRP-BCAM/sBCAM were developed by mold pouring process, a series of human skull physical phantom was made by 3D printer, and the numerical calibration
method of computational phantom was also established. In addition, several intercomparison of whole body counter measurement and internal dose assessment in
China was organized by China Institute for radiation protection. A series of in vivo
monitoring technology with completely independent intellectual property rights has
been researched and developed through nearly ten years of team work.
At the same time, it is noted that the accuracy, consistency of measurement results
and standardization of dose assessment program need to be further improved according to the results of intercomparison. Therefore, it is still a very worthy research

43

Invited speakers

04

issue to establish an economical and coordinated procedure for the individual monitoring of internal exposure due to intakes of radionuclides and improve the reliability
and accuracy of monitoring results by combining various bioassay such as in vivo
monitoring, in vitro analysis and aerosol sample analysis.

44

05 Invited speakers
General Features of PHITS and its Applications to
Cosmic-Ray Dosimetry
Tatsuhiko SATO
Japan Atomic Energy Agency, Tokai, Ibaraki, JapanThe Particle and Heavy Ion Transport code System (PHITS) is a general-purpose
Monte Carlo radiation transport simulation code that can analyze the motion ofradiations in any material using the nuclear/atomic reaction models and libraries.
It has been used in more than 60 countries for various applications such as radiation shielding design, medical physics, and cosmic-ray researches. In thispresentation, the general features of the PHITS code together with its applications to cosmic-ray dosimetry including the developments of PARMA (PHITSbased Analytical Radiation Model in the Atmosphere) and WASAVIES (WArning
System for AVIation Exposure to Solar energetic particle) will be introduced.

45

Invited speakers

06

Development and Applications of Multi-scale
Dosimetry with Chinese Reference Phantom Library
Rui Qiu1,* , Zhen Wu2, Ankang Hu1, Xiyu Luo1, Jiahao Wang1, Ziyi Hu1,
Shuchang Yan1, Wanyi Zhou1, Junli Li1
Department of Engineering Physics, Tsinghua University, Beijing, China1,
Nuctech Company Limited, Beijing, China2
Purpose: Latest development and applications of multi-scale dosimetry with the
Chinese reference phantoms are introduced in this paper. Computational phantoms are fundamental tools of radiation dosimetry. A library of Chinese reference
phantoms is required for the dose estimation for the Chinese population. The
multi-scale Chinese phantom library has been developed, including whole-bodyphantoms, detailed organ models and nucleus models to cover the scales of
millimeters, micrometers and nanometers. Previous work has established theframework of the multi-scale phantom library, including Chinses adult reference
phantoms and pediatric phantoms on the millimeter scale, the detailed skeleton
model, eye model, breast model and respiratory tract model on the micrometerscale and atomic DNA model on the nanometer scale. A series of Monte Carlosimulation methods have been developed for multiscale dosimetry with the phantom library.
Methods & Results: Recently, the multi-scale phantom library and multi-scale
dosimetry have been developed to expand the range, improve the efficiency andprecision, and had various applications. A body-size-dependent series of Chinese adult phantoms was established based on mesh-type phantoms. The series
covers more than 90% Chinese population, containing 42 phantoms with a height
range of 145-185 cm and weights corresponding to healthy, overweight andobese. The differences of organ dose coefficients between a phantom and areference phantom vary by up to 40% for the same height and up to 400% for
the whole phantoms. The dose conversion coefficients of digital breast tomosynthesis were calculated based on the detailed breast model, considering a
range of breast and exposure parameters. For the aspect of efficiency, a photonelectron coupled transport Monte Carlo simulation code was developed with GPU
acceleration for the fast dose estimation of X-ray radiography. The speedup is
up to ~200. A GPU-accelerated code has been developed to accelerate particle
transport in polygon-mesh and tetrahedron-mesh geometries. Preliminary tests
showed that the code improves the speed by about 30 times. For the aspect of
personalized dosimetry, a personalized dose estimation method for nuclear medicine was developed. The nuclear medicine images are constructed by a deeplearning algorithm to obtain personalized radiopharmaceutical distribution. The
personalized phantom is constructed based on clinical images and a personalized46

06 Invited speakers
pharmacokinetic model is extracted based on clinical data. These tools provide apipeline for personalized and precise internal dose assessment for radionuclide
therapy. For the nanodosimetry study, the atomic DNA model and the trackstructure Monte Carlo code NASIC have been utilized to explore the mechanismof FLASH effect from the perspective of radiation physics and chemistry. This
study analyzed the oxygen depletion hypothesis and proposed a novel theory of
FLASH effect mechanism, “radical recombination and antioxidants” hypothesis.
Conclusion: The multiscale dosimetry with the multi-scale Chinese reference
phantom library provides efficient and valuable tools for dose estimation for the
Chinese population. It also supports the exploration of the frontiers in radiation
dosimetry and radiobiology.

47

Invited speakers

07

Multi-scale modeling in Radiation Therapy –Bridging the gaps between Physics and Biology
Jan Schuemann
Massachusetts General Hospital & Harvard Medical School, Boston, MA, USA

Understanding the effects of radiation therapy is essential to deliver optimaltreatments. Over the years, many models have been proposed to capture theseeffects, from simple macroscopic models based on dose, linear energy transfer(LET) or a combination of these, to more detailed models including effects at thecell scale. Modeling approaches offer a strong tool to understand mechanismsof complex processes, allowing us to probe correlations that are not accessible
with experimental techniques.
The Monte Carlo method is an exceptionally useful tool to describe the propagation of radiation through matter. My lab uses the Monte Carlo method to investigate the effects of radiation on tissue at multiple scales. We try to understand how initial physics processes at the DNA scale can result in cell damage
and repair. We then try to apply our models to investigate outcome of different
therapies with the goal to find the best treatments design for, for example, external beam therapy, FLASH therapy or radiopharmaceutical therapy.
Here, I present an overview of our activities, trying to model clinical to cell level
effects and the models and tools we developed along the way.

48

08 Invited speakers
Dose Coefficients of the ICRP for external exposures –Current Development and Status

Nina Petoussi-Henss
ICRP Committee 2
Major efforts within ICRP* Committee 2 are the development of reference dosecoefficients for exposure to external environmental and occupational radiation
fields and those for inhalation and ingestion of radionuclide. These efforts build
upon changes in radiation and tissue weighting factors and definition of effective
dose (Publication 103), radionuclide decay schemes (Publication 107), development of reference computational phantoms of the adult (Publication 110) andpaediatric male and female (Publication 143) and development of skeletal dosimetry. In this presentation, the overall computational framework for external
dose coefficients will be reviewed. Moreover, the development of dose coefficients for occupational settings will be discussed as well as the development ofdose coefficients to members of the general public – adult, adolescents, children
and infants– due to external environmental exposures.
For many years, the ICRP has produced reference dose coefficients of the patient for common diagnostic nuclear medicine procedures. However, ICRP has
not provided so far reference dose coefficients of the patient for X-ray imaging
procedures. Recently the ICRP embarked on a major effort to produce coefficients between absorbed or equivalent dose to organs at risk and measurable
quantities commonly used in X-ray imaging procedures – for radiographic, CTand fluoroscopic examinations. Several radiation protection organisations aroundthe world have simulated diagnostic radiology examinations on computational
phantoms using Monte Carlo techniques. These simulated dose coefficientsmight not represent the current state of the art and moreover have been obtainedusing a variety of anthropomorphic phantoms and codes. ICRP has decided toharmonize this process by determining dose coefficients of the patient using the
ICRP methodology, which is worldwide recognized and adopted by national andinternational radiation protection organizations. The results of this on-going effort will be reviewed and the limitations will be discussed.

*International Commission on Radiological Protection

49

Invited speakers

09Radiological Protection in Nuclear Medicine: ICRP Activities andGuidelines for PET/CT and Radionuclide Therapy
Keon Wook KANG
Seoul National University, Seoul, Korea
The International Commission on Radiological Protection (ICRP) emphasizes theimportance of radiological protection in nuclear medicine, particularly in the areasof positron emission tomography (PET), hybrid imaging, and therapy with radiopharmaceuticals. For PET and PET/CT facilities, proper shielding, automation,and facility planning are crucial to protect patients, staff, and the public. Justification and optimization of imaging protocols should consider the characteristicsof the imaging technologies and establish national diagnostic reference levels(DRLs) for both PET and CT components.
Occupational doses in PET should be monitored closely, and appropriate
shielding devices and optimized working practices must be adopted to minimize
exposure. Radiological protection education and training are essential for eachmember of the medical imaging team, ensuring quality assurance program goals
are met.
In radionuclide therapy, the goal is to optimize the relationship between tumor
control probability and potential complications in normal organs and tissues. Absorbed dose calculations are usually performed prior to therapy using trace-labeled diagnostic administrations, or retrospectively after each therapeutic administration.
Accident prevention in radiation therapy should be integrated into facility design,
equipment, and administration procedures. Minimizing staff exposures includes
equipment design, proper shielding, handling of sources, personal protective
equipment, and education and training. The decision to hold or release a patient
after radiopharmaceutical therapy should consider potential radiation dose to the
public and carers, and specific radiological protection guidance should be provided accordingly.

50

Abstracts - Oral List
No.

Topic

Title

Authors

Affiliation

O101

Evaluation of Shielding
Material against Galactic
Cosmic Radiation: Monte
Carlo Simulations with
ICRP-110 Reference
Phantom

Li Xiaoyu1,Ye
Zirui1,Chen Zhi1

O102

Re s e a rc h o n Co m p lex
Source Term Inversion andRadiation Field Reconstruction Method Basedon Deep Neural Networks

Yisheng HAO1,Zhen
WU2,Yanheng
PU1,Rui QIU1,Hui
ZHANG1,Junli LI1

1

School of Nuclear Science and Technology,
University of Science and Technology of China,
Hefei, China

1

Department of Engineering Physics, Tsinghua
University, China|2Department of Engineering
Physics, Tsinghua University, China|Nuctech
Company Limited, China

O103

A Preliminary Study on a
GPU Accelerated Monte
Carlo Dose Calculation
Method for BNCT

Yuxin Wang ,Shijun Li1,Jin Wu1,
Zirui Ye1,
Li Tao1,Xi Pei2,
Xie George Xu3

1School of Nuclear Science and Technology,
University of Science and Technology of China,
China|2Anhui Wisdom Technologies, Inc., China|3School of Nuclear Science and Technology,
University of Science and Technology of China,
China|Department of Radiation Oncology, The
First Affiliated Hospital of University of Scienceand Technology of China, China

O104

Radiation Exposure Assessment for Outdoor
Large-Structure Tests in
Industrial Radiography

Junik CHO,Euidam
KIM,Wonhyeong
LEE,
Yoonsun CHUNG

1
Department of Nuclear Engineering, Hanyang
University, Korea

Study on induced radioactivity of different materials
in the proton therapy facility

Yue Yu1,Zhi Chen

O106

Three-dimensional dynamic γ radiation fieldcalculation method for
digital simulation of radiation protection

Haijing Jin1,
Hua Li1,Yuan
Zhao1,Zeyu Long2,
Jiafu Wei2,Liye Liu1

O107

The Research on Neutron
Shielding Performance and
The Secondary γ Dose
Simulation of Tungsten
Boride Material

Yi Han1,
Xiaomiao
Chi1,Faguo
Chen1,Guodong Li1

O108

M o n t e C a r lo A n a ly s i sConsidering Temperature
Change of Metal Alloy
Materials in the In-Core
Irradiation Test

Hae SunJeong1,Kee Nam
Choo1,Sung Jae
Park1,
Ye-Eun Na1,
Junesic Park1,
Yoon Taeg Shin1,
Chul Yong Lee1,
Seong Woo Yang1

O109

H i g h f re q u e n c y p u l s eemission characteristics
and application of carbon
nanotube cold cathode Xray source

Sheng Lai1,
Yunpeng Liu1,
Junxu Mu1,
Kai Miao1,
Xiaobin Tang1

O201

Experimental evaluationof inter-fractional dosimetric changes in vaginal
cuff brachytherapy using
in-house designed PMMA-agar phantom

Saerom Sung1,
Bo-Wi Cheon1,
Chul Hee Min1,
Hyemi Kim2,
Sei Hwan You2,
Hyun Joon Choi2

1

O105

#1 ‌
Radiation
Transport
andShielding#2
Radiation
Dosimetry

1

51

School of Nuclear Science and Technology,
University of Science and Technology of China,
Hefei, China

1

China Institute for Radiation Protection, China
Shanxi Provincial Key Laboratory for Radiation
Safety and Protection, China

21China Institute for radiation protection , Taiyuan, China

1

HANARO Utilization Division, KAERI, Korea

1
Department of Nuclear Science and Technology, Nanjing University of Aeronautics and Astronautics, Nanjing Jiangsu, China

1

Department of Radiation Convergence Engineering, Yonsei University, Wonju, Korea|2Department of Radiation Oncology, Wonju Severance Christian Hospital, Yonsei University
Wonju College of Medicine, Wonju, Korea

Abstracts - Oral List
No.

Topic

Title

Authors

Affiliation

O202

Impact of mesh-type Korean reference phantoms
on dose calculations forcomputed tomography(CT) scans

Yumi LEE1,
Ji Won CHOI1,
Haegin HAN2,
Choonsik LEE3,
Yeon Soo YEOM1

1
Department of Radiation Convergence Engineering, School of Yonsei University, Korea|2Department of Nuclear Engineering, Schoolof Hanyang University, Korea|3Division of Cancer Epidemiology and Genetics, National Cancer
Institute, National Institute of Health, U.S.A.

O203

Geometric Considerationsin L-band EPR Tooth Dosimetry Using Radiation
Treatment Plans

Jong In Park1,
Jeonghun Oh2,
Chang Uk Koo2,
In Jung Kim1,
Kwon Choi2,Sung-Joon Ye1

1
Ionizing Metrology Group, Korea ResearchInstitute of Standards and Science, South Korea| 2Department of Applied Bioengineering,
Graduate School of Convergence Science andTechnology, Seoul National University, South
Korea
1

Jin Wu ,Shijun Li1,
Yuxin Wang1,
Qiang Li3,
Xie George Xu2

School of Nuclear Science and Technology,University of Science and Technology ofChina, Hefei, China |2Technology Development
Department, Anhui Wisdom Technology Company Limited, Hefei, China |School of Nuclear
Science and Technology,University of Scienceand Technology of China, Hefei, China |3Biomedical center,Institute of Modern Physics,
Chinese Academy of Science, Lanzhou, China

Zirui Ye1,
Xiaoyu Li1,
Zhi Chen1,
Jingnan Guo2,
Xie George Xu3

1School of Nuclear Science and Technology,
University of Science and Technology of China,
Hefei, China|2School of Earth and Space Sciences, University of Science and Technology ofChina, Hefei, China|3School of Nuclear Scienceand Technology, University of Science andTechnology of China, Hefei, China|Departmentof Radiation Oncology, The First Affiliated Hospital of USTC, Hefei, China

1

GPU-Based Fast Monte
Carlo Code for Carbon Ion
Therapy

O204

Preliminary Phantom andMonte Carlo Space Radiation Dosimetric Studies
for Lunar Missions

O205

#2
Radiation
Dosimetry1

A script-based automaticcommissioning method of
virtual source model forindependent dose verification module

Bo Cheng1,
Yuan Xu2,
Shijun Li1,
Qiang Ren3,Xi Pei3,
Kuo Men2,
Jianrong Dai2,
Xie. George Xu4

School of Nuclear Science and Technology,
University of Science and Technology of China,
China|2Department of Radiation Oncology, National Cancer Center|National Clinical ResearchCenter for Cancer|Cancer Hospital, Chinese
Academy of Medical Sciences and Peking
Union Medical College, China|3Anhui Wisdom
Technology Company Limited, China.|4Schoolof Nuclear Science and Technology, Universityof Science and Technology of China, China|Department of Radiation Oncology, The First Affiliated Hospital of USTC, University of Scienceand Technology of China, China

O207

Age-dependent Iodine-131
S values Based on ICRP
Reference Pediatric Phantoms

Soo Min Lee1,
Yumi Lee1,Ji Won
Choi1,Chul Hee
Min1,Hyun Joon
Choi2,Chansoo
Choi3,Yeon Soo
Yeom1

1
Department of Radiation Convergence Engineering, Yonsei University, Korea|2Departmentof Radiation Oncology, Yonsei University Wonju
College of Medicine, Korea| 3J Crayton Pruitt
Family Department of Biomedical Engineering,
University of Florida, FL, USA

O208

CPU-GPU Coupling Monte
Carlo Acceleration Method
for Dose Calculation of
Mesh-type Phantom

Shuchang Yan1,
Rui Qiu1,
Xiyu Luo1,
zhen Wu1,
Hui Zhang1,Junli Li1

O209

Personalized Cumulated
Activity Distribution Calculation using Biokinetic
Compartmental Model in
PET Imaging

He Yutao1,
Chen Zhi1

O206

52

1

Department of Engineering Physics, Tsinghua
University, China

1

School of Nuclear Science and Technology,
University of Science and Technology of China,
ChinaAbstracts - Oral List
No.

Topic

Title

Authors

Analysis of Occupationaldoses in Korean nuclear
power plants over the
past ten years

Chang Ju Song1,
Tea Young Kong1

O211

Development of Virtual
Dose Monitoring System
for Interventional Radiology (VIR)

Gahee Son1,
Haegin Han1,
Sungho Moon1,
Bangho Shin1,
Hyeonil Kim1,
Suhyeon Kim1,
Soon Young Song2,
Chan Hyeong Kim1

O212

Dose coefficients for soil
contaminations produced
using mesh-type Korean reference phantoms(MRKPs)Ji Won Choi1,
Yumi Lee1,
Bangho Shin2,
Chansoo Choi3,
Yeon Soo Yeom1

O213

Patient-Specific Dose
Assessment for Nuclear
Medicine Therapy Basedon 177Lu-PSMA

Ziyi Hu1,Ruiyao
Ma1,Rui Qiu1,
Ankang Hu1,Zhen
Wu2,Hui Zhang1,
Junli Li1

O214

Establishment and first
application of 4D dose
calculation in China

Zhaoxing Liu1,
Ri Zhao2,Na Liu1,
Jing Zhang2,
Runcheng Liang2,
Xin Liu1

Establishment of High
Dose Rate Pulsed γ Radiation Facility

Zhihui Tang

O216

Establishment of radiation
biological dose estimation
curve based on automatic
analysis of dicentric chromosomes

Li Shufang1,
Qin Xiujun2,
Yin Jingjing3,
Xu Wenli4,
Zhang Yajuan5, Yue
Juan6,
Li Jianguo7,*

O217

A performance study of anew personal neutron dosimeter based on SSNTD

Yan Jiao1,Xiaofeng
Wei1,Qinjian Cao1

O210

O215

#2
Radiation
Dosimetry

Affiliation
1

School of Chosun University, Korea

1

Department of Nuclear Engineering, College ofEngineering, Hanyang University, Korea| 2Department of Radiology, Hanyang University
Hospital, Korea

1

Department of Radiation Convergence Engineering, Yonsei University, Korea |2Departmentof Nuclear Engineering, Hanyang University,
Korea|3J Crayton Pruitt Family Department of
Biomedical Engineering, University of Florida,
U.S.A.
1

Department of Engineering Physics, Tsinghua
University, Beijing, China|2Nuctech Company
Limited, Beijing, China

1

China Institute for Radiation Protection, Taiyuan, China|2Shanxi Provincial Key Laboratory forTranslational Nuclear Medicine and Precision
Protection, Taiyuan, China

1

China Institute for Radiation Protection, China

1

Department of Radiology and Environmental
Medicine, China Institute for Radiation Protection, China

1

China Institute for Radiation Protection, China

1

O218

Fast and Accurate GPUBased Monte Carol Independent Dose Verification
Platform for Proton Therapy

Shijun Li1,Li Tao1,
Xi Pei2,
XieGeorge Xu3

School of Nuclear Science and Technology,
University of Science and Technology of China,
Hefei, China|2Anhui Wisdom Technologies, Inc.,
Hefei, China|3School of Nuclear Science andTechnology, University of Science and Technology of China, Hefei, China|Department of Radiation Oncology, The First Affiliated Hospital of
University of Science and Technology of China,
Hefei, China
1

O219

Demonstration of Realtime Monte Carlo Person-specific Organ Dose
Computing involving Automatic Segmentation of
10,000 Patients

Zirui Ye1,Yang
Lu2,Haoran
Zheng3,Bei Yao3,
Xi Pei2,
Xie George Xu4

53School of Nuclear Science and Technology,
University of Science and Technology of China,
Hefei, China|2Anhui Wisdom Technologies, Inc.
Hefei, China|3Health Center, The First Affiliated Hospital of USTC, Hefei, China|4School of
Nuclear Science and Technology, University of
Science and Technology of China, Hefei, China|Department of Radiation Oncology, The First
Affiliated Hospital of USTC, Hefei, China

Abstracts - Oral List
No.

Topic

O301

Title

AuthorsD e v e l o p m e n t o f Tw o
Ty p e s o f L i g h t w e i g h t
Portable Neutron Survey
Meter

Takashi Nakamura1,
Tomoya Nunomiya1,
Seini Yamamura1

Affiliation
1

Fuji Electric Co. Ltd.|2Tohoku University

Xin Wang1,Yubing
Xu1,Yuzhu Pan1,
Omolola Esther
Fayemi,Qing Li,
Wei Lei1School of Electronic Science and Engineering,
Joint International Research Laboratory of Information Display and Visualization, Southeast
University, Nanjing, China| 2 Department ofChemistry, School of Mathematics and Physical Sciences Faculty of Natural and Agricultural
Sciences North-West University, Mafikeng
Campus, Private Bag X2046, Mmabatho 2735,
South Africa

O303

Low-noise and High sensitivity direct-conversion
X-ray detectors based on
perovskite single-crystal
heterojunctions

Yuzhu Pan1,
Xin Wang1,Yubin
Xu1,Omolola Esther
Fayemi2,Qing Li1,
Ying Zhu3,Wei Lei1

1
Joint International Research Laboratory ofInformation Display and Visualization, Schoolof Electronic Science and Engineering, Southeast University, Nanjing, China|2Department ofChemistry, School of Mathematics and PhysicalSciences Faculty of Natural and Agricultural
Sciences North-West University, Mafikeng
Campus, Private Bag X2046, Mmabatho 2735,
South Africa|3E-Xray Electronic Co. Ltd., Suzhou, China

O304

Application of Unfoldingcode MEALU to Neutron
Spectrum Measurement
using Multi-cylindrical
Bonner Counter

Tetsuo Iguchi1,
Shigetaka Maeda2

1
Nagoya University, Japan|2Japan Atomic Energy Agency, Japan

Development of an inhouse PMT voltage divider
for the TDCR counting
systemAgung Agusbudiman1,
Kyoung Beom
Lee,Jong Man
Lee,Sang Hoon
Hwang,Byoung
Chul Kim

Department of Precision Measurement,
University of Science and Technology (UST),
Korea| 2Radioactivity Metrology Team, Korea
Research Institute of Standards and Science(KRISS), Korea

O306

Study on FPD for highenergy X-ray CT for fueldebris screening

Akira Mikami1,
Yuki Mitsuya1,
Hiroyuki Takahashi1

1

O307

Feasibility Verification of
Deep Learning Based 4
π Field of View Collimator-less Imaging System
Using Voxelated Single
Volume GAGG(Ce) scintillator: A Monte-Carlosimulation

Ajin Jo1,
Dongmyeong Hong2,
Wonho Lee3

1
Health Science Research Center, Korea University, Korea|2Department of Health and Safety
Convergence Science, Korea University, Korea|3
School of Health and Environmental Science,
Korea University, Korea

Minato Kanda1,
Kenji Shimazoe2,
Yusuke Tamura3,
Hanwool Woo4,
Mizuki Uenomachi5,
Fumihiko Ishida6,
Jun Kawarabayashi7,
Ken'ichi Tsuchiya8,
Kei Kamada9,
Hideki Tomita1

1
Graduate school of Engineering, Nagoya University, Japan|2Graduate school of Engineering, The University of Tokyo, Japan|3Graduate
school of Engineering, Tohoku University,
Japan|4Faculty of Engineering, Kogakuin University, Japan| 5Center for the Promotion ofInterdisciplinary Education and Research, Kyoto
University, Japan|6Department of Electrical andControl Systems Engineering, National Instituteof Technology, Toyama College, Japan|7Facultyof Science and Engineering, Tokyo City University, Japan|8Second Department of Forensic
Science, National Research Institute of Police
Science, Japan|9New Industry Creation Hatchery Center, Tohoku University, Japan

Detection of gamma-ray
photons with halide leadperovskite single crystals
for high energy resolution

O302

O305

O308

#3
Radiation
Detectionand Sensor
Technology

Demonstration of Gamma
Source Finding in the Field
based on Fusion Databetween 4π Gamma Imaging and Simultaneous
Localization and Mapping

54

1

University of Tokyo, JapanAbstracts - Oral List
No.

Topic

Title

Research on Detection
Efficiency Calibration by
GEANT4 Simulation fort h e H P G e D e te c to r i n
HTR-PM

O309

Authors
Rui Nie1,Yu
Wang1,Ziling
Zhou1,Qian Ma2,Xi Chen2,
Zihao Fan3,
Juncheng Liang3,
Feng Xie1,
Jianzhu Cao1

Affiliation
1

Institute of Nuclear and New Energy Technology, Collaborative Innovation Center ofAdvanced Nuclear Energy Technology, Key
Laboratory of Advanced Reactor Engineeringand Safety of Ministry of Education, Tsinghua
University, Beijing, China|2Huaneng Shandong
Shidao Bay Nuclear Power Company LTD,
Rongcheng，Shandong , China|3National Institute of Metrology, Beijing, China
1High-energy photon
counting based on Sed o p e d M A P b I 3 s i n g lecrystals

Yubing Xu ,Xin
Wang1,
Shunjie Chai1,
Yuzhu Pan1,
Omolola Fayemi2,
Qing Li1,Ying Zhu3,
Wei Lei1

School of Electronic Science and Engineering,
Joint International Research Laboratory of Information Display and Visualization, Southeast
University, Nanjing, China| 2 Department ofChemistry, School of Mathematics and Physical Sciences Faculty of Natural and Agricultural
Sciences North-West University, Mafikeng
Campus, Private Bag X2046, Mmabatho 2735,
South Africa|3E-Xray Electronic Co. Ltd., Suzhou 215131, P. R. China

O311

Ultra-dense Ceramic
Scintillators for Positron
Emission Tomography

Sun Il Kwon1,Yimin
Wang2,Minseok Yi3,
Urmila Shirwadkar2,
Steven Onorato2,
Junghune Nam2,
Jae Sung Lee3,
Jarek Glodo2,
Kanai Shah2,
Simon Cherry11Department of Biomedical Engineering, University of California Davis, Davis, CA 95616,
USA| 2Radiation Monitoring Devices, Inc., 44
Hunt St., Watertown, MA 02472, USA|3Interdisciplinary Program in Bioengineering, College of Engineering, Seoul National University,
Seoul, 03080, South Korea

O312

Development of Glass Gas
Electron Multiplier based
Beta Ray Imaging System

Moh HAMDAN1,
Yuki MITSUYA2,
Kenji SHIMAZOE1,
Hiroyuki TAKAHASHI1

1

O310

#3
Radiation
Detectionand Sensor
Technology

O313

Dosimetric Study of a
Novel Perovskite Scintillator for High-Energy
Radiotherapy

Ning Gao1,
Zongming Huang2,
Jieping Zhou4,
Yihan Zhang3,
Xi Pei5,
Zhengguo Xiao2,
Xie George Xu1

O314

Background Reductionin PET using Quantum
Entanglement Property ofAnnihilation Gamma Rays

Donghwan Kim1,
Kenji Shimazoe1,
Hiroyuki Takahashi1

O315

Ra d i o n u c l i d e a n a ly s i sUsing PHITS and Neutron
Irradiation in HANARO
Concrete Sample

Shin Dong
Hyuk1,Kim Ji Seok1,
Kown Jin Hyung1

O316

Development of Prototype
Backscatter X-ray Imaging System for Security
Screening

Geunyoung AN1,
Junsung PARK1,
Hee SEO1,Byeong-no LEE2

O317

Development of Background Simulation Technique for Radiation Portal
Monitors at Airports andPorts in South Korea

Hayoung SIM1,
Hyosang
CHOI1,Hee
SEO1,Hyeonjun
CHOI2,Haneul
KIM2,Jaeho JANG2

1

Department of Nuclear Engineering and Management, Graduate School of Engineering, The
University of Tokyo, Japan|2Institute of Engineering Innovation, School of Engineering, The
University of Tokyo, Japan

1
School of Nuclear Science and Technology,
University of Science and Technology of China, Hefei, China|2Hefei National Laboratory for
Physical Science at the Microscale, Departmentof Physics, CAS Key Laboratory of Strongly-Coupled Quantum Matter Physics, Universityof Science and Technology of China, Hefei, China|3Institute of Energy, Hefei Comprehensive
National Science Center, Hefei, China|4Department of Radiation Oncology, The First AffiliatedHospital of University of Science and Technology of China, Hefei, China|5Technology Development Department, Anhui Wisdom Technology
Company Limited, Hefei, China
1

Department of Nuclear Engineering and Management, Graduate School of Engineering, The
University of Tokyo, Japan

1

55

Department of Nuclear Engineering, Hanyang
University, Seoul, Korea|2HANARO Utilization
Division,Korea Atomic Energy Research Instittute, Daejeon, Korea
1

Jeonbuk National University, Jeonju, Korea|2Korea Atomic Energy Research Institute,Jeongeup, Korea
1

Department of Applied Plasma and Quantum
Beam Engineering, Jeonbuk National University, Korea|2Department of Radiation Monitoringat Ports of Entry, Korea Institute of Nuclear
Safety, Korea

Abstracts - Oral List
No.

Topic

Title

Authors

Performance Estimationof PVT Scintillator-based
Seafood Monitoring System for Screening of Radioactive Materials

Hyosang CHOI1,
Hee SEO1

1

O319

Radiation Hardness Assurance Testing of a 40nm Microprocessor Applied in Nuclear Industry

Runcheng Liang1,
Faguo Chen1,
Rong Guo1,
Zhirui Zheng1,
Ri Zhao1,Jing
Zhang1,Zhaoxing
Liu1,Xing Liu1

1

O320

Scintillation Properties
of Cr-doped α-Al2O3
Crystals for the Decommissioning Monitoring
SystemShunsuke Kurosawa1,Akihiro
Yamaji2,Daisuke
Matsukura2,
Akira Yoshikawa2

O321

Research on the performance of SiC thermal
neutron detector basedon metal-oxide-semiconductor structure

Lei Ren1,YunchengHan1,Jiacheng
Zhang1,Rui
Zhang1,Houjun
He1,Xiaoyu Wang1

O322

Proposal of Method for
E st i m a t i o n o f G a m m a
Source Direction based on
Self-shielding Profile in a
Multipixel Array Detector

Atsushi Mukai ,Minato Kanda2,
Hideki Tomita2

Fuel debris mapping tool
using an Unmanned Underwater Vehicle to grasp
fuel debris submerged in
Fukushima Daiichi Nuclear
Power Plant

So Kamada1,
Tetsuichi Kishishita2,
Takahiro Oyama2,
Kenichi Terashima3,
Keisuke Okumura2,
Kazuya Nishimura3

Fabrication of Transition
Edge Sensors with faster
response time

Takeshi Jodoi,
Yuki Mitsuya,
Ryan Smith,Amin Choghadi,
Hiroyuki Takahashi

Department of Engineering, The University ofTokyo, Japan

O325

Muography-based method of subsidence risk assessment for undergroundtunnel construction

Naoya okamoto1,
Shoichiro Kawase1,
Ayumu Okuda1,
Hiroto Shiba1,
Yukinobu
Watanabe1

1

O326

Waveform discrimination
by fitting derivative ofsynchronized ideal normalized curves, dSINC fit

Martin Andersson1,
Tatsuo Torii2,
Susumu Ryufuku1,
Ryohei Kurosawa1,
Hiroko Kido1

1
Visible Information Center, Inc., Tokai-mura,
Japan2Institute of Environmental Radioactivity,
Fukushima University, Japan

O327

Development of Dual
Energy X-ray Security
Scanner for Improved Organic Material Detection:
A Monte Carlo Study

Junsung Park,
Geunyoung An,
Hee Seo

O328

Characteristic evaluationof gamma-ray spectrometer using an unfolding
method with refined response functions

Masateru Hayashi1,
Makoto Sasano1,
Tetsushi Azuma2,
Taisuke Makita1,
Masakazu
Nakanishi2,
Hiroshi Nishizawa3

O318

O323

O324

#3
Radiation
Detectionand Sensor
Technology

Affiliation

Jeonbuk National University, Jeonju, Korea

China Institute for Radiation Protection, China

1

Tohoku University & Osaka University, Japan|Tohoku University, Japan

2

1

Hefei Institute of Physical Science, Chinese
Academy of Sciences, China

1

1

56

Department of Energy Engineering, Graduate School of Engineering, Nagoya University,
Japan|2Department of Aplied Energy, Graduate
School of Engineering, Nagoya University, Japan1

N a t i o n a l M a r i t i m e Re s e a rc h I n s t i t u t e ,Japan|2High Energy Accelerator Research Organization, Japan|3Japan Atomic Energy Agency,
Japan

1

IGSGS, Kyushu University, Japan1

Department of Applied Plasma and Quantum
Beam Engineering, Jeonbuk National University, Republic of Korea

1

Advanced Technology R&D Center, Mitsubishi Electric Corporation, Japan| 2Energy Systems Center, Mitsubishi Electric Corporation,
Japan|3Department of Applied Nuclear Technology, Fukui University of Technology, JapanAbstracts - Oral List
No.

Title

Authors

Affiliation

O329

Multi-Class Radioisotope
Identification using Convolutional Neural Networks with Two-Dimensional Input Data

Yong Hyun
Kim1,Yong Kyun
Kim1,Yang Han
Cheol,Kihong Pak,
Jae Chang Kim,Jae
Young Jeong1
Department of Nuclear Engineering, Hanyang
University, Seoul, Korea

O330

Pile-up Separation and
Pulse Shape Discrimination for an Organic Scintillation Detector using
Multi-task Learning

Byoungil Jeon1,
Junhyeok
Kim2,Myungkook
Moon1

1
Korea Atomic Energy Research Institute, Korea|2Korea Advanced Institute of Science andTechnology, Korea

Application of a novel
scintillation gel-based
detector for 3D dosimetry
in radiotherapy

Hua Li1,Liang
He3,Haijing Jin1,
Xuewen Yan1,
Hui Zhang2,
Deyuan Li2,Liye Liu1

China Institute for Radiation Protection,
China| 2Shanxi Provincial Key Laboratory forTranslational Nuclear Medicine and Precision
Protection, China|3Department of Engineering
Physics, Tsinghua University, China

Building and performance
test of X-ray communication system based on
grid-controlled modulated X-ray tube and LYSO-SiPM X-ray receiverJunxu Mu1,Yunpeng
Liu1,Xiaobin Tang1

1
Department of Nuclear Science and Technology, Nanjing University of Aeronautics and Astronautics, China

O401

Characterization of doseand time-dependent radiation effects on humanskin by multiomics andhistological analyses

Won-Suk Song2,
Jae-Seung Lee2,
Jun Woo Lim2,
JiEung Kim1,
Ji-Hyeon
Park2,Sang Hyoun
Choi3,Dongchan
Jang1,
Il Won Kim2,
Jae Hyun Jeong2,
Yun-Gon Kim2

1
Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Scienceand Technology (KAIST), Korea|2Department ofChemical Engineering, Soongsil University, Korea|3Research Team of Radiological Physics andEngineering, Korea Institute of Radiological andMedical Sciences (KIRAMS), Korea

O402

Development of a measuring method of cosmic-ray muon momentum
distribution using drift
chambers

Naoto Nakagami1,
Satoko Kamei1,
Shoichiro Kawase1,
Akira Sato2,Satomi
Nakatani1,Yukinobu
Watanabe1

Determination of Ra-226
in Biological Samples via α-spectrometry Using Ra223 as Yield Tracer

Ping XU 1,
Guobing YU 2,
Huiyun XUE 1,
Huiping TANG 1,
Shiquan CAO1,Chen
ZHI1,Yujia KANG1

O404

Numerical analysis anddiscussion on the implementation of airborne
radioactive emission standards of nuclear power
plant radiation protection
standard (GB6249-2011)

Fu Shen

O405

The content level andparticle size distribution of
210Po in Taiyuan environmental aerosol

Pengxiang Li,
Xiaona Ren,
Xuyuan Ma,Feng
Zhang,Ruijun
Wang,Kaite Lu

O406

Study on the Influence of
Large Natural Ventilation
Cooling Towers on Atmospheric Diffusion at Offshore NPPs

Junfang ZHANG1

O331

Topic

#3
Radiation
Detectionand Sensor
Technology

O332

O403

#4
Environmental
Radiation
Measurement andAssessment

57

1

1

Kyushu University, Fukuoka, Japan| 2Osaka
University, Osaka, Japan

1

University of Science and Technology of China,
Hefei 230027, China| 2Anhui Radiation Environmental Supervision Station, Hefei 230071,
China

1

China Institute for Radiation Protection,China

1

China Institute for Radiation Protection, Taiyuan, 030006, China

1

Department of Nuclear Environmental Science,
China Institute for Radiation Protection, China

Abstracts - Oral List
No.

Title

Authors

O407

Concentration behavior
a n d ra d i a t i o n d o s e o f
137Cs in zebrafish

Shaofei
CAO,Qingliang
FENG,
Wenjing GONG

O408

Preliminary Study on Separation and Concentration for Large Volume of
Krypton by Improved Gas
ChromatographXUYUAN MA1,
JING ZHANG,RUIJUN WANG,
PENGXIANG LI,ZHOU LI

O409

Analysis of Interference
between Tritium and Carbon -14 Measurement

Li BAO,Bing LIAN,
Youkun YANG,
Chen GUO,Hailan
YANG,Xuyuan MA

O410

Study on Optimal Allocation Method of Dose Constraint Value in Nuclear
Facilities

Hailong CHEN1,
Bing LIAN1

The Graded Approach in
the Radiation Environment
Management of NORM in
China

Wang Yan1,
Lian Bing1,
Gu Zhijie1

Models and Parameterstudies for Retrospective
Evaluation of Nuclear Facilities

Jing Kang1,
Bing LIAN1,
Feifei WU1

O413

Application of Deconvolution Method to Alpha-ray
Energy Spectrum to
Improve Performance ofRadiation Airborne Particulate Monitor

Makoto Sasano1,
Masateru Hayashi1,
Yusuke Yanagawa4,
Masatoshi Kai4,
Yasushi Nakano4,
Yoshitsugu Osawa4,
Shunsuke
Kurosawa2,
Yuki Morishita3

O414

CFD Simulation of Resuspension Release of Radioactive Aerosol Particles
from Contaminated Sites

Jia-chen
CHEN, Hai-long
CHEN,Bing LIAN ,Yan WANG

O415

Parameter sensitivity
analysis of different ecological environments fora tritium environmental
transfer model

Junyu CHENG1,
Rentai YAO1

O501

Radiological Assessmentof Loss of Coolant Accident from Shin Kori Nuclear Power Plant

Moses Oboo,
Jouyoul Kim

Radiological Risk Assessment of Safety Exit Sign
Containing Tritium.

Yeijin BANG1,
Jimin SHIN1,
Sunlee SHIN1,
Hee SEO1,
Byeonghyeon
PARK2,
Sangmin LEE2

Department of Applied Plasma and Quantum
Beam Engineering, Jeonbuk National University, Korea|2Department of Radiaion Regulation,
Korea Institue of Nuclear Safety, Korea

Radiation Dose Assessment of Thorium-containing Gas Mantle for
P re d i c t a b le E x p o s u re
Scenarios

Jimin SHIN1,
Yeijin BANG1,
Sunlee SHIN1,
Hee SEO1,
Byeonghyeon
PARK2,
Sangmin LEE2

1
Department of Applied Plasma and Quantum
Beam Engineering, Jeonbuk National University, Korea|2Department of Radiation Regulation,
Korea Institute of Nuclear Safety, Korea

O411

O412

O502

O503

Topic

#4
Environmental
Radiation
Measurement andAssessment#5Radiological
Risk
Management

Affiliation
1

Institute of Nuclear Environmental Science,
China Institute for Radiation Protection, China

1

China Institute for Radiation Protection, China

1

Key Laboratory of Radiation Environment &Health of the Ministry of Ecology and Environment, China Institute for radiation proctection,
China

1

Key Laboratory of Radiation Environment &Health of the Ministry of Ecology and Environment, China Institute for Radiation Protection,
China

1

China Institute for Radiation Protection, Tai Yuan, China

1

China Institute for Radiation Protection, Key
Laboratory of Radiation Environment & Healthof the Ministry of Ecology and Environment,
Taiyuan, 030006, China

1

Mitsubishi Electric Corp., Japan|2Tohoku University, Japan|3Japan Atomic Energy Agency,
Japan| 4Mitsubishi Electric Plant Engineering
Corp., Japan

1

China Institute for Radiation Protection, China

1

China Institute for Radiation Protection, China

158

Kepco International Nuclear Graduate School(KINGS), Department of Nuclear Power Plant
Engineering, South Korea

1Abstracts - Oral List
No.

Topic

Title

Authors

Assessment of Source
Terms and Potential Doses Due to Steam Generator Tube Rupture of
VVER-1200 at El Dabaa
Nuclear Power Plant in
Egypt

waad saleh1,
Juyoul kim1

Thermal Safety Assessment of Transport Containers for Radioactive
Materials

Changwu Wang1,
Yuhang Zhang1,
Yiren Lian1,
Lei Chen1,
Zhipeng Wang1,
Qian Sun1

1

China Institute for Radiation Protection, China

O506

Discussion of “Applicationof knowledge management in radioactive waste
management”

enbang liu1,
xiliang guo1,
chao gao1,
yanjiao wang1,
xu han1,yahui xi1,
xiaojun yan11China Institute For Radiation Protetion, China

O601

Tutorial Development for
Digital Twin Simulationin Scenario-based Linear Accelerator Daily QA
Training

Wonhyeong Lee1,
Joengeun Kim1,
Yoonsun Chung1

1
Department of Nuclear Engineering, Hanyang
University, Korea

Report on the Association
for Radiation Science andApplications as a Collaboration of Academia andIndustry in Japan

Tetsuo Iguchi1,
Hiroyuki Takahashi2

1
Nagoya University, Japan|2The University ofTokyo, Japan

O701

Validation and applicationof migration diffusionmode in nuclear accident
consequence evaluation
systemMINGHUA LYU1,RENTAI YAO1,JUNFANG ZHANG1,DUOXIN ZHAO1,SHA HUANG1

O702

Investigation on Decision-Making Criteria for
Public Protective Actions
in Nuclear Emergencies

Wi-Ho Ha1,
Chanki Lee1,
Ilje Cho1,
Won Tae Hwang1,
Hyun Ki Kim1

A visual-based positioning
and guidance approachfor nuclear emergency
response manipulator

Qi Lv1,Xiaoyu
Zhao1,Yuyao Tang1,
Yi Liu1,
Yapeng Yang1

1

O704

Machine Learning-based
Denoising Technique for
Gamma Emission Tomography to Inspect Partial
Defects of Spent Nuclear
Fuel

Hyung-Joo Choi1,
Yoon Soo Chung1,
Hyun Joon Choi2,
Kyunghoon Cho3,
Hakjae Lee3,
Hyung Cheol Lee4,
Yong Hyun Chung1,
Chul Hee Min1

1
Department of Radiation Convergence Engineering, Yonsei University, Republic of Korea|2Department of Radiation Oncology, Wonju
College of Medicine, Republic of Korea|3ARALE
Laboratory Co. Ltd, Republic of Korea|4Nuclear
Material Analysis Team, Korea Insitute of Nuclear Nonproliferation and Control, Daejeon,
Republic of Korea

O801

Factor Analysis of In-Situ
O b j e c t Co u n t i n g S y s tem(ISOCS) for Uranium
Quantification in Radioactive Waste Drums

Jihyun Ahn1,
Seungmin Lee2,
Hee Seo1

1School of Jeonbuk National University, Korea|2Department of Korea Institute of Nuclear
Nonproliferation and Control, Korea

Development of AI-based
Anomaly Detection Modelfor Nuclear Waste Data

Jinhyung Kwon,
Young-su Kim,
Jiseok Kim,
Hanjin Kim,
Donghyeok Shin,Yongkyun Kim

O504

O505

O602

O703

O802

#5
Radiological
Risk
Management#6
Education,
Training
and Policyin Radiation
Safety#7
Radiological
Emergency
Planning
andPreparedness#8
Radioactive
Waste andCurrent
Radiological
Issues59

Affiliation

1

Department of radiation safety and environmental protection, Kepco nuclear graduate
school, South korea

1

Dapartment of Nuclear environmental science
research, China Institute for Radiation Protection, China

1

Department of Nuclear Emergency Preparedness, Korea Atomic Energy Research Institute,
Korea

China Institute for Radiation Protection, China

1

Korea Atomic Energy Research Institute(KAERI), Daejeon, Korea|2Department of Nuclear Engineering, Hanyang University, Seoul,
Korea

Abstracts - Oral List
No.

Title

Authors

O803

Research on the key problems of the new smelting
and purification process ofmetal radioactive pollutants

Kewei Jiang

1

O804

The Study of Mechanical wear principle anddynamics of graphite
spheres in HTR-PM hightemperature gas cooled
reactorLixiao Guo

1

O805

Development of Microwave Drying Technology
for Radioactive Wet Waste
Treatment in Drum

Meng Zihe1,
Gao Chao1,
Jia Meilan1,
Liang Dong1

O806

Development and verification of portable inc i n e ra t i o n t e c h n o lo g y
for low-level radioactivewaste

Chu Haoran1,
Xu Wei1,
Zheng Bowen1,
Zhang yu1,
Ruan Jiasheng1

O807

Research on Si-Al baseGeopolymers Formulationof Simulated Radioactive
Fluoride Liquid Wastes

Li Zhang,
Xiliang Guo

O808

The Study on Microwave
Drum Drying for Radioactive Wet Waste

Gao Chao1,
Jia Meilan1,
Meng Zihe1,
Yan Xiaojun1

A synthesized bifunctional
strongly basic anion exchange resin for ReO4-/
TcO4- removal

Yujia Kang1,
Ping Xu1,
Huiping Tang1,
Shiquan Cao1,
Huiyun Xue1,
Zhi Chen1

Research and applicationo f o n - s i t e c a l i b ra t i o ntechnology for a radioactive waste drum NDA
inspection system

Hengguan YI1,
Zhigang LI1,
Zhihui TANG1,
Zhenglin HUANG1,
Yingjing WEI1,
Qinjian CAO1

O811

Activation status of metallic components in the
PET-cyclotron revealed bydestructive and non-destructive analysis

Go YOSHIDA1,
Hiroshi
MATSUMURA1,
Akihiro TOYODA1,
Hajime NAKAMURA1,
Kazuyoshi
MASUMOTO1,Taichi MIURA1,
Iwane KOUMURA2,
Hiroyuki UNO2,
Masahiko KOZAKI3,
Ayumi
YAMAMOTO3

O812

Efficient Adsorption of
Sr2+ from High Level Liquid Waste by 4A Zeolite

xiang qin1,
chao gao1,
xiliang guo1,
xia tao1

O813

Dose estimation for radioactive metal waste handling workers using doseconversion factors of ICRP
Publications 119, 141, and
EPA FGR No. 11

Jin Ho SON1,Tae Young KONG1

O809

O810

Topic

#8
Radioactive
Waste andCurrent
Radiological
Issues60

Affiliation

China Institute for Radiation Protection, China

China Institute for Radiation Protection, China

1

The Department of Radioactive Waste Management, China Institue for Radiation Protection, China

1
China Institute for Radiation Protection, Taiyuan, China

1

China Institute for Radiation Protection, China

1Radioactive Waste Management Department,
China Institute for Radiation Protection, China

1

School of Nuclear Science and Technology,
University of Science and Technology of China,
Hefei, China

1

Shanxi Provincial Key Laboratory for Radiation
Safety and Protection, Taiyuan, China & Key
Laboratory of Radiation Protection of CNNC,
Taiyuan, China & China Institute for Radiation
Protection, Taiyuan, China

1High Energy Accelerator Research Organization (KEK), Tsukuba-Ibaraki, Japan|2Sumitomo
Heavy Industries, Ltd., Tokyo, Japan|3SHI-ATEX
Co., Ltd., Seika-Kyoto, Japan|4Tokyo Metropolitan Geriatric Medical Center, Tokyo, Japan

1

China Institute for Radiation Protection, China

1

Department of Nuclear Engineering, Chosun
University, Gwangju, Korea

Abstracts - Oral List
No.

Title

Authors

O814

Speciation simulation andsorption mechanism of
238Pu in waste repository
groundwater

CHEN Chao,
ZHANG Aiming,LUO Zhiping

O815

Gaseous radioactive
iodine capture of microporous molecular sieves
supported silver: BEA,
MFI and CHA

Zexiang Chen1,
Jie Yu1,Xin Li1,
Jianlu Pei1,
Shuwei Liang1,
Yongguo Li1

The improvement in theiodine capture property ofsilica-based aerogel byintroducing three-dimensionally ordered macroporous

Sen Chang1,
Yongguo Li1,
Jianli Chen1,
Jie Yu1,
Jianrong Hou1,
Dangui Qiu1,
Qun Liu1

Thermodynamic Analysis
of Effect of Waste Resin
Irradiation Gas on Mineral
Composition of Ordinary
Portland Cement

HAN Xu1,
GAO Chao,YAN Xiaojun,
XI Yahui,
QIN Xiang,
LIU Enbang,
GUO Xiliang

O816

O817

Topic

#8
Radioactive
Waste andCurrent
Radiological
IssuesAffiliation
1

61

Department of Nuclear Environmental Science, China Institute for Radiation Protection,
China|2Department of Radiation Safety, China
Institute of Atomic Energy, China

1

Key Laboratory of Nuclear Air Purification
Technology, China Institute for Radiation Protection, China

1

Department of Environment Engineering andTechnology, China Institute for Radiation Protection, China

1
The Department of Radioactive Waste Management, China Institute for Radiation Protection, China

Abstracts - Oral

O1 - 01#1 Radiation Transport and Shielding

Evaluation of Shielding Material against Galactic Cosmic
Radiation: Monte Carlo Simulations with ICRP-110 Reference
Phantom

Li Xiaoyu1, Ye Zirui1, Chen Zhi1,*
School of Nuclear Science and Technology, University of Science andTechnology of China, Hefei, China 1
Email: zchen@ustc.edu.cn
Keywords: shielding material, Monte Carlo, Galactic Cosmic Radiation, phantom
Purpose: The exposure to galactic cosmic radiation (GCR) is a major healthconcern for astronauts. Due to the complexity of GCR and the self-shielding
characteristics of human body, effective shielding materials must strike a balance
between increasing and decreasing energy deposition and biological effects of
the radiation. Additionally, the use of radiation weighting factors in spaceradiation protection is not recommended, yet many studies still choose thismethod as the evaluation criterion. Instead, the organ dose equivalent H_T
defined by the quality factor is recommended by the NCRP and ICRP.
Methods: In terms of those problems, this study aims to develop a method toeffectively evaluate the GCR shielding effect of space materials. Monte Carlosimulations, coupled with ICRP-110 reference phantom, were conducted using
the Geant4 code to assess the performance of different thicknesses of aluminum(0, 5, 20 g/cm²) as shields against GCR. This simulation chose the BadhwarO’Neill 2020 (BON 2020) model as the isotropic ambient field. The Particles
between Z = 1 and Z = 28 were included in all transport calculations. The
geometric part mainly consisted of a spherical aluminum shield and a phantom
located in the center. Three quantities were calculated as evaluative criteria:
absorbed dose, equivalent dose, and the flux of the particle through the phantom.
Absorbed dose and equivalent dose estimate the effectiveness of shielding
materials with the existing evaluation system, whereas the flux of the particle
through the phantom may offer a new perspective to the evaluation.
Results: This study has demonstrated that the performance of different
thicknesses of aluminum as GCR shields is similar, with the equivalent dosevarying by no more than 8%. Furthermore, the study revealed that GCR doses did
not show large organ dependence other than the skin. The 20 g/cm² shieldthickness was found to reduce the equivalent dose to the skin by 55% compared62

#1 Radiation Transport and Shielding

O1 - 02#1 Radiation Transport and Shielding

Research on Complex Source Term Inversion and Radiation Field
Reconstruction Method Based on Deep Neural Networks

Yisheng HAO1, Zhen WU2, Yanheng PU1, Rui QIU1, Hui ZHANG1, Junli LI1,*
Department of Engineering Physics, Tsinghua University/China1, Department ofEngineering Physics, Tsinghua University/China2, Nuctech Company
Limited/China2
Email: lijunli@mail.tsinghua.edu.cn
Keywords: Deep Neural Network，Source Term Inversion，Radiation Field
Reconstruction，Monte Carlo Method
To accurately evaluate the radiation protection for on-site operators of nuclear
installations and correspondingly reduce the radiation dose that operators are
exposed to through the lifecycle of nuclear facilities, it is necessary to accurately
evaluate the distribution of radiation source terms and parameters in theoperating environment. However, due to limitations in the size and quantity ofsuch measuring devices, close-packed methods cannot be used to detect the
distribution of these parameters during actual operations. Thus, determining how
to calculate the distribution of radiation parameters and source terms in the entire
exposure space from individual sampling points has become a key issue.
In actual radiation environments, the relationship among radiation measurement
values at each sampling location, including the distribution of source terms andradiation parameters in the space, is highly nonlinear. This makes the radiationdifficult to describe with a definite mathematical model. Artificial intelligence
methods, particularly neural networks, can often effectively solve such nonlinear
problems and may thus be an effective tool for solving these problems.
For the distribution of radiation parameters, we use discrete sampling point
gamma flux recordings as input data and deep neural network methods to train
and test evaluation of the gamma radiation fields under different source term
distributions, based on three-dimensional geometric examples. Verification
results show that the neural network method can efficiently predict the distributionof radiation parameters, with the maximum value of the average percentagechange error on the test set being 3.66%, which is less than the error range(±10%) of the given input sampling points.

63

Abstracts - Oral

O1 - 02

Regarding the inversion problem of source term distribution, we propose a deepneural network (SDICNN) model to obtain distribution information of complex
fixed source terms from the radiation parameters of individual sampling points inspace. The SDICNN is divided into two parts: the fully connected network part
mainly extracts measurement parameter information from the sampling points,
and the convolutional neural network part mainly completes the fine inversion of
the distribution of the source items. We test source term inversion using common
geometric scenes and find that compared to the traditional fully connected neuralnetwork, the SDICNN network can extract the two-dimensional features of thesource term distribution, converge faster, and produce better prediction results.
Additionally, we study the influence of factors such as shielding and the numberof sampling points on the inversion effect of the source term.

64

#1 Radiation Transport and Shielding

O1 - 03#1 Radiation Transport and Shielding

A Preliminary Study on a GPU Accelerated Monte Carlo Dose
Calculation Method for BNCT

Yuxin Wang1, Shijun Li1, Jin Wu1, Zirui Ye1, Li Tao1, Xi Pei2, Xie George Xu3,*
School of Nuclear Science and Technology, University of Science andTechnology of China/China1, Anhui Wisdom Technologies, Inc./ China2, Schoolof Nuclear Science and Technology, University of Science and Technology ofChina/China3, Department of Radiation Oncology, The First Affiliated Hospitalof University of Science and Technology of China/China3
Email: xgxu@ustc.edu.cn
Keywords: GPU Monte Carlo, Dose calculation, BNCT, Geant4
Purpose: At present, boron neutron capture therapy (BNCT) generally uses Monte
Carlo method to ensure the accuracy of dose calculation. The Monte Carlo
method demands a high number of particles and prolonged computational time
to minimize statistical errors, which substantially curtails the clinicalimplementation of BNCT. A new method of GPU accelerated Monte Carlo Method
was developed to improve the speed of neutron and secondary particles
transport, which can be used in the dose calculation of the BNCT treatment
planning system.
Method: Our program is implemented under the Compute Unified Device
Architecture (CUDA) platform to exploit the full potential of the GPU's extended
thread-level parallelism and superior efficiency. The elastic scattering, nonelastic scattering, and capture reaction of neutron over whole energy range inthe BNCT process were simulated. The secondary particles such as photon andproton were simulated by the GPU Accelerated Monte Carlo software ARCHER.
We ascertained the accuracy of our computation by comparing it to simulations
performed with Geant4. A water phantom model was simulated.
Results: In the simulation of a water phantom, our program has a good agreementwith Geant4 in simulating physical processes. The dose calculations time of ourcode was within 3 minutes with a single Nvidia GTX 1060 6G card. While Geant4
took about 85 minutes running on computation platform with the Intel(R) Xeon(R)Gold 6348 CPU with 100 threads. Through the simulation of photons and protons,
we discovered that the GPU can increase the computational speed of the Monte
Carlo program by approximately 100 times.
Conclusion: We have conducted preliminary research on the Monte Carlo dose
calculation simulation method accelerated by GPU in BNCT. Based on the results

65Abstracts - Oral

O1 - 03of our study, utilizing the GPU-accelerated Monte Carlo method to simulate thetransport of neutron and other secondary particles (e.g. photons) is a feasible
approach for accurate dose calculation in BNCT while maintaining computational
efficiency. Moving forward, we will continue to refine our program to further
enhance its GPU acceleration capabilities and integrate it into a treatment
planning system DeepPlan.

66

#1 Radiation Transport and Shielding

O1 - 04#1 Radiation Transport and Shielding

Radiation Exposure Assessment for Outdoor Large-Structure
Tests in Industrial Radiography

Junik CHO, Euidam KIM, Wonhyeong LEE, Yoonsun CHUNG*
Department of Nuclear Engineering/Hanyang University/Korea1
Email: ychung@hanyang.ac.kr
Keywords: Radiation assessment, Occupational exposure, Radiation exposure,
Industrial radiography
Purpose: Industrial radiography is a non-destructive testing method that uses
radiation to inspect products for defects, but it can also pose a serious risk toworkers' health due to radiation exposure. Workers in the field of industrial radiography have one of the highest annual average radiation doses among radiation-related workers, and radiation accidents often occur.
According to the Korean Institute of Nuclear Safety (KINS) report, outdoor largestructure tests involve a high workload and consequently a high exposure risk.
Also, the demand for outdoor large-structure tests is increasing due to the recent
revival of the domestic shipbuilding industry. There are cases in which collimators
cannot be used when testing large outdoor large-structures, and even if a collimator is used the amount of radiation exposure may vary depending on the location where workers are waiting during exposure.
In this study, radiation exposure was evaluated for workers for outdoor largestructure tests in the field of industrial radiography.
Methods: Industrial radiography of ship block welding points was selected as arepresentative outdoor large-structure test. We assumed that the ship block wasmade of EH47 metal which is commonly used in modern ships and a tungsten
collimator with a 60° beam angle.
In the field of industrial radiography, Ir-192 is widely used as a radiation source,
and 0.397keV, the average air kerma-weighted energy, was selected as the representative energy. The activity of a radiation source for outdoor large-structure
tests depends on the characteristics of the object, such as material properties
and thickness. Considering these characteristics, Ir-192 sources with an activity
level of under 20 Ci are commonly used in industrial radiography for ship inspection.
The Particle and Heavy Ion Transport code System (PHITS) code was used as aradiation exposure assessment tool. The PHITS-derived particle fluence was thenused with the dose conversion coefficients specified in the International Com-67

Abstracts - Oral

O1 - 04

mission on Radiological Protection (ICRP) 60 report to calculate the effective
dose. To calculate the annual exposure dose, the calculated effective dose isderived considering the working hours. Through a survey of radiation workers, it
was found that the outdoor large-structure test work accounted for about 25%of the total radiography work. Considering work proportion from the survey andworking hours for outdoor large-structure test presented in the KINS report, theworking hours were assumed to be 160.5 hours in a year. From this, the annual
exposure dose from outdoor large-structure tests was derived. Also, a water
phantom was placed where workers could potentially wait during exposure. The
locations where the water phantom is located include outside the ship (A), inside
the ship where the worker waits in the inspected bulkhead (B), and inside theship where the worker waits in the different bulkhead (at locations 1 m (C) and 2
m (D) away from the adjacent wall of bulkhead). In addition, when a collimator is
not used, we assumed that the worker staying in the inspected bulkhead waits at
a distance of 20 m from the source. For all other cases except this last one, theworkers were assumed to be 10 m away from the source.
Results: When using an Ir-192 source with an activity level of 20 Ci for radiography and using a collimator, radiation exposure was derived high in the order of
B, A, C, and D (approximately 1.946, 0.329, 0.024, and 0.016 mSv/yr, respectively). Assuming a collimator is not used for the entire working hour, the radiation
exposure at locations A, B, C, and D was 1.05, 21.9, 2.78, and 4.51 mSv/yr,
respectively, which is much higher than when a collimator is used due to insufficient shielding. Particularly, location B had the highest radiation exposure since
there was nothing to shield the workers. When the collimator was not used, theexposure at location B was more than 10 times higher than when the collimator
was used, even though the water phantom was 20 m away from the source. The
risk of exposure could be higher because the worker needs to approach about
10 m to operate the remote controller.
Conclusion: In this study, we calculated the annual exposure dose of workers in
outdoor large-structure tests depending on the use of a collimator. If a collimatorcannot be used for a special case, workers at location B where the radiation
exposure is relatively high are recommended by using a remote device or additional shielding. Furthermore, it was confirmed that moving to other bulkheads
resulted in a significant reduction in radiation exposure. It is anticipated that thefindings of this study will contribute to improved safety culture and increased
awareness of radiation safety among workers.

68

#1 Radiation Transport and Shielding

O1 - 05#1 Radiation Transport and Shielding

Study on induced radioactivity of different materials in the proton
therapy facility

Yue Yu1, Zhi Chen*
School of Nuclear Science and Technology, University of Science andTechnology of China, Hefei, China1
Email: zchen@ustc.edu.cn
Keywords: Monte Carlo Simulation, FLUKA, proton therapy facility, induced
radioactivity
Proton therapy facilities have proliferated in recent years, bringing benefits topeople and potential radiation hazards and risks. Induced radioactivity in proton
therapy facility components is essential for occupational personnel's radiation
safety and maintenance, repair, and decommissioning activities. So far, many
researchers have extensively reported and explored the topic of induced
radioactivity caused by proton radiotherapy in the literature. They all believe thatinduced radioactivity generated at proton accelerator hot spots poses a newchallenge to radiation safety. Meanwhile, induced radioactivity in patients is also
an important issue that has been ignored for a long time and deserves further
study and discussion. Therefore, this paper will study the induced radioactivity
generated by patients and surrounding materials during proton therapy, assess
its impact and make recommendations to reduce the induced radiation hazardsof patients and staff.Considering the maturity and convenience, the Monte Carlocode FLUKA was adopted and RADDECAY, IRRPROFI, DCYTIMES, RESNUCLE,
and DCYSCORE input cards were used to calculate activation. The beams ofproton of 130 MeV and 250 MeV are incident on a rectangular water phantom of
30 cm× 30 cm× 70 cm. To conservatively estimate the induced radioactivity ofthe patient and other materials, a simplified estimation model was established.
Results: (1) Materials involved in the treatment room include tissue (patient),copper (nozzle), iron (local shield), stainless steel (patient bed support), andpolyethylene (patient bed). The dominating nuclides produced in the patientswere C-11, N-13, and O-15, and that of copper were Cu-62 and Cu-64. Mn52, Fe-53, Mn-52m, and Fe-53m were produced in iron. Whereas the types of
nuclides produced in stainless steel were more, including Al-28, V-48, Mn-52,
Mn-52m, Mn-54, Mn-56, Fe-53, Fe-53m, Co-56, etc. The dominant nuclide inpolyethylene is C-11. (2) Patients themselves were the major contributors to theresidual dose rate in the treatment room after a single treatment, but their69

Abstracts - Oral

O1 - 05

importance decreased over time because most radionuclides produced inpatients were short half-life ones. (3) Different levels of activation resulted fromdifferent irradiation energies, with the residual dose rate after two minutes ofirradiation at 130 MeV being about half that of 250 MeV. Depending on the energy
required by different types of tumors, the time for staff to enter the treatment
room for positioning or escorting can be determined. It takes only one minute for
130 MeV and five minutes for 250 MeV.

70

#1 Radiation Transport and Shielding

O1 - 06#1 Radiation Transport and Shielding

Three-dimensional dynamic γ radiation field calculation methodfor digital simulation of radiation protection

Haijing Jin1, Hua Li1,*, Yuan Zhao1, Zeyu Long2, Jiafu Wei2, Liye Liu1
China Institute for Radiation Protection, China 1, Shanxi Provincial Key
Laboratory for Radiation Safety and Protection, China2
Email: lihua_7559@126.com
Keywords: γ radiation field, digital simulation, point kernel integration theory,
weight discretization, source activity inversion
Digital simulation technology has brought an easy-to-use and perceptible
approach for radiation protection, which is essential for the optimization ofradiation protection at nuclear facility. However, the unknown of radiation source
at nuclear facility and the low efficiency of current radiation detection equipment
resulted in huge difficulties for accurate simulation of the dynamic γ radiation
field. Such problems limited the feasibility and performance of the visualization
simulation technology application in on-site radiation protection.
In this paper, based on the point kernel integration theory, the buildup factorsthat affects calculation accuracy was studied. Based on Geant4 platform andusing the finite medium model, the buildup factor database of γ radiation under
two combinations of common materials (water, iron, lead, and concrete) was
calculated. The difference between the calculation results and the data obtainedby empirical formulas and other methods was compared and analyzed. So was
the difference between the calculation results and the corresponding data using
the infinite medium model. And the data were verified with specific arithmetic
examples. The new buildup factors reduced the deviation between the point
kernel integration calculation results and the Geant4 simulation results by about
15%, which greatly improved the calculation accuracy of the point kernel
integration software.
This paper also studied the discrete method of non-point sources that affects
the efficiency of point kernel integration calculation. Through a detailed
mathematical derivation of the line source discretization, a new physical quantity“discrete size n” was defined for weight discretization. Based on this physicalquantity, the effect of weight discretization for different source types wascalculated and analyzed in detail, and the quantification index of weight
discretization was given for subregions. For a cubic source case, the variableweight discretization method was used. Compared with the constant71

Abstracts - Oral

O1 - 06

discretization method, a calculation efficiency improvement of about 75 times
was achieved and the calculation error of both discretization methods was
guaranteed to be within 5%.
Based on the obtained results, this paper also established a dynamic γ radiation
field calculation method based on source activity inversion. Firstly, use measured
data of the radiation field, combined with the 3D geometric model and other priori
data of the radiation source (such as location, quantity and nuclide type and ratio,etc.), to calculate the source activity in reverse. Then use the derived source
activity as input to calculate the 3D radiation field data of the site in forward
simulation. For the non-uniform source activity distribution, the Bayesian
information criterion was used to select the optimal source activity distribution
model, and the maximum likelihood expectation maximization algorithm was used
for iterative solution. The established method was verified for a room in a nuclear
power plant. The results showed that the average deviation between thereconstructed γ radiation field and the measurement value was within ±8%, while
the percentage of measurement points within two standard deviations (2σ) was
above 90%. Consequently, the reconstruction accuracy met the needs of on-site
calculation analysis.
The relevant research of this paper will serve the digital simulation of radiation
protection, focusing on the real-time calculation of dynamic γ radiation field data,
realizing dynamic data-driven, and effectively improving the effect of theradiation protection digital platform for practical application in nuclear facilitysites.
Acknowledgement
This research work was supported by the Talented Young Scholar Project (No.29)
of China National Nuclear Corporation.
Reference
1. Vetter K, Barnowski R, Cates JW, Haefner A, Joshi THY, Pavlovsky R, et al.
Advances in Nuclear Radiation Sensing: Enabling 3-D Gamma-Ray Vision.
Sensors. 2019 Jan;19(11):2541. 2. Szőke I, Louka MN, Bryntesen TR, Edvardsen
ST, Bratteli J. Comprehensive support for nuclear decommissioning based on 3D
simulation and advanced user interface technologies. Journal of Nuclear Scienceand Technology. 2015 Mar 4;52(3):371–87.

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#1 Radiation Transport and Shielding

O1 - 07#1 Radiation Transport and Shielding

The Research on Neutron Shielding Performance and The
Secondary γ Dose Simulation of Tungsten Boride Material

Yi Han1,*, Xiaomiao Chi1, Faguo Chen1, Guodong Li1
China Institute for radiation protection , Taiyuan, China1
Email: heaven0china@163.com
Keywords: GEANT4, W-B series compounds, Neutron absorption, Secondary
γ-ray
In view of the demand for neutron shielding materials in terms of weight reduction,body reduction and high temperature resistance in the development process ofvehicle/ship-borne small modular of nuclear reactors, small modular of space
reactors, fusion reactors, etc. It is necessary to realize neutron absorption andsecondary γ Analysis of ray shielding performance of new shielding materials.
In this paper, the GEANT4 program is used to analyze the neutron absorption andsecondary γ-ray shielding properties of W-B series compounds which mainly
include W2B, WB, WB2, W2B5 and WB4. The simulation calculation is mainly
carried out for the thickness of the shielding material from 0 to 2 cm and theincident neutron energy from thermal neutron to 20 MeV.
The results show that: 1) W-B series compounds mainly act on the absorption
and shielding of neutron energy from thermal neutrons to 10-2 MeV; When the
thickness is constant, the neutron absorption performance of W2B5 is the best;
When the mass is constant, WB4 has the best neutron absorption performance.
2) From the neutron macroscopic removal cross section and material density ofdifferent materials, compared with traditional neutron absorbing materials, W-B
series compounds have more significant advantages in low-energy neutron
absorption. Taking thermal neutrons as an example, the neutron macroscopic
removal cross section of W2B5 material is about 8.67 times that of B203 material
and 40.59 times that of PB202 shielding material.3) the contribution of thesecondary γ dose to the total dose decrease with the increase of the incident
neutron energy, and the contribution of the secondary γ dose to the total dose
increase with the increase of the thickness of the shielding material.
Our study provides a reference for the realization of the optimal design of neutron
shielding devices for clarifying the application advantages and application
scenarios of W-B series compounds, and has practical engineering guidancevalue.

73

Abstracts - Oral

O1 - 08#1 Radiation Transport and Shielding

Monte Carlo Analysis Considering Temperature Change of Metal
Alloy Materials in the In-Core Irradiation Test
Hae Sun Jeong1,*, Kee Nam Choo1, Sung Jae Park1, Ye-Eun Na1, Junesic Park1,
Yoon Taeg Shin1, Chul Yong Lee1, Seong Woo Yang1
HANARO Utilization Division, KAERI, Korea1
Email: haesunin@kaeri.re.kr
Keywords: Monte Carlo Neutronics Transport Analysis, Doppler Broadening Effect,Heat-Temperature Iteration Solution, Temperature Reactivity, Neutron Spectrum
The HANARO research reactor in KAERI has proven nuclear and mechanical integrity for various reactor-related materials through in-core irradiation tests.
Meanwhile, the Monte Carlo neutronics analyses for various irradiation tests havegenerally been performed based at the fixed room temperature condition. In thisstudy, the steady-state transport problems along with temperature variation by
the reactor operation were solved to evaluate whether the room temperaturebased calculations produce reasonably acceptable results. The MCNP6 Monte
Carlo code was used for radiation transport analysis, and the 1-D GENGTC code
was used for temperature analysis of the target materials that are directly affectedby the internal neutron and gamma-ray heat generation. By generating the
ENDF-based nuclear cross-section libraries for the individually updated target
temperature, a series of correlation calculations were iterated with the converged
values between the two codes. The irradiation test was assumed to be performed
using the hexagonal CT (Central Thimble) irradiation hole located in the center ofthe core, which shows the highest neutron flux in the reactor core. The irradiation
capsules are composed of five different sections vertically loaded in the irradiation hole. The target specimen is an iron-based alloy with a density of 7.80 g/cm3.
The target areas are divided into four or six equal parts and the target materials
are located by symmetrically rotated in the horizontal direction. In this study, it isassumed that the reactor power remains a constant of 30 MW and only the target
temperature is set to be the main parameter to be evaluated for effective analysis
on Doppler effects. Along with temperature changes by the reactor operation,
the heat generation by neutrons and gamma-rays were calculated for the iterative
correlation calculations, and the temperature reactivity and neutron spectra on
the target materials were analyzed. The results of this study will be used to determine the irradiation characteristics in the actual reactor operation and to prove
the safety of the irradiation test. The detailed results will be presented at theconference.
Acknowledgement
This work was supported by the Korea government (MSIT) (1711173832).
74

#1 Radiation Transport and Shielding

O1 - 09#1 Radiation Transport and Shielding

High frequency pulse emission characteristics and application ofcarbon nanotube cold cathode X-ray source

Sheng Lai1, Yunpeng Liu1, Junxu Mu1, Kai Miao1, Xiaobin Tang1,*
Department of Nuclear Science and Technology, Nanjing University of Aeronautics and Astronautics, Nanjing Jiangsu, China1
Email: tangxiaobin@nuaa.edu.cn
Keywords: Radiation transport characteristics, CNT X-ray source, High-frequency pulse emission, X-ray communication
Carbon nanotube (CNT) cold cathode X-ray source has the advantages of controllable emission and instantaneous response, and has great application potential in high-frequency pulse application.
In order to reveal the radiation transport characteristics of high-frequency pulse
X-ray based on CNT cold cathode X-ray source, the cathode current emission
characteristics and the relationship between pulse X-ray and voltage, frequency
and pulse amplitude are researched by experiment, exploring the time characteristics under the high-frequency pulse emission condition, and carry out the
X-ray communication (XCOM) data transmission experiment.
The radiation transport characteristics results show that the turn-on electric fieldof cathode is 0.1 mA@2.5 V/um. The threshold electric field is 1 mA@3.81 V/um.
When the grid voltage is higher than 470 V, the amplitude of pulse X-ray waveform reaches peak, and the shape presents a pseudo-square wave. The time
characteristics results of X-ray high-frequency radiation transport show thatwhen the grid voltage increases from 600 V to 650 V, the X-ray emission delay
time drops sharply from 102 ns to 75 ns, and the relative pulse broadening of Xray waveform tends to be saturated, about 350 ns (± 20 ns), indicating that thecorresponding theoretical limit frequency of the X-ray radiation transport in thisstudy is about 2.86 MHz (1/350 ns). In the XCOM data transmission experiment,
the CNT cold cathode X-ray source is used as a communication signal sourcewith 1.05 MHz pulse transmission frequency. At different pulse radiation transport
frequencies, the pulse X-ray waveform and the voltage waveform of loading data
signal have a good consistency.
This work reveals the pulse radiation transport characteristics of CNT cold cathode X-ray source and its influencing factors, and realizes the high-frequency
pulse emission of CNT cold cathode X-ray source, which lays a foundation forthe development and application of CNT cold cathode X-ray source in XCOM
and pulse application.
75

Abstracts - Oral

O2 - 01#2 Radiation Dosimetry

Experimental evaluation of inter-fractional dosimetric changes invaginal cuff brachytherapy using in-house designed PMMA-agar
phantom

Saerom Sung1, Bo-Wi Cheon1, Chul Hee Min1, Hyemi Kim2, Sei Hwan You2,
Hyun Joon Choi2,*
Department of Radiation Convergence Engineering, Yonsei University, Wonju,
Korea1, Department of Radiation Oncology, Wonju Severance Christian
Hospital, Yonsei University Wonju College of Medicine, Wonju, Korea2
Email: hjchoi1@yonsei.ac.kr
Keywords: vaginal cuff, brachytherapy, PMMA-agar phantom, inter-fractionalchange, dosimetric evaluation
Vaginal-cuff brachytherapy is a common adjuvant treatment for women withintermediate-high risk of endometrial or uterine cervical cancers. The mostcommonly used applicators are single-channel vaginal cylinders and twosymmetric ovoid applicators, with the distance between them determined by thetherapist and the size of the vagina. A greater distance between the ovoids leads
to a lower dose at the vaginal apex point, which can be problematic. Therefore,
in this study, we experimentally evaluate dosimetric changes by inter-fractional
patient-setup changes using in-house designed PMMA-agar phantom andpropose the use of three-ovoid applicators to improve insufficient dose
distribution.
The PMMA phantom consists of a box-shaped frame measuring 30×20×10 cm³,with multiple holes for inserting A1SL and Farmer ionization chambers, Gammex
solid rods, and a cylindrical agar phantom with a diameter and length of 10 cm.
To create the agar phantom, 950 ml of distilled water and 500 g of agar powder
were mixed in a 95:5 ratio with an ovoid applicator inserted. The BEBIG
MultiSource® High Dose Rate brachytherapy unit with a cobalt-60 source wasused to deliver the treatment beam for various applicator setups in the agar
phantom, and dosimetric evaluation was performed using ion chambers atseveral points in the phantom.
The density of the agar phantom was sufficient to stably maintain the position of
the inserted applicators. Thin A1SL chamber or tandem applicators could be
easily inserted at desired positions without the need for fixation, providing
flexibility in changing the experimental setup. The cylinder applicator inserted into76

#2 Radiation DosimetryO2 - 01

the PMMA-agar phantom delivered doses of 1.2 Gy and 1.5 Gy at the virtual
bladder and rectum points, respectively, and 4.4 Gy at the apex. Insertion of twoovoid and three-ovoid applicators resulted in dose evaluations of approximately
50% and 101%, respectively, relative to this apex point dose.
In this study, we experimentally evaluated dosimetric changes caused by interfractional patient setup changes for vaginal cuff brachytherapy using the inhouse designed PMMA-agar phantom. Using this phantom, we experimentallyconfirmed the feasibility of the three-ovoid applicators consisting of two ovoids
and tandem with multiple holding discs to compensate for dose coverage at thevaginal apex cold spot due to the greater distance between the two ovoids.
Acknowledgement
This research was supported by the Basic Science Research Program throughthe National Research Foundation of Korea (NRF) funded by the Ministry of
Education (NRF-2021R1I1A1A0105987512), the Korea Institute of Energy
Technology Evaluation and Planning (KETEP), and the Ministry of Trade, Industry& Energy (MOTIE) of the Republic of Korea (No. G032579813).

77

Abstracts - Oral

O2 - 02#2 Radiation Dosimetry

Impact of mesh-type Korean reference phantoms on dose
calculations for computed tomography (CT) scans

Yumi LEE1, Ji Won CHOI1, Haegin HAN2, Choonsik LEE3, Yeon Soo YEOM1,*
Department of Radiation Convergence Engineering, School of Yonsei
University/Korea1, Department of Nuclear Engineering, School of Hanyang
University/Korea2, Division of Cancer Epidemiology and Genetics, National
Cancer Institute, National Institute of Health/U.S.A.3
Email: ysyeom@yonsei.ac.kr
Keywords: Computed Tomography, Organ dose, Monte Carlo transport, Koreanmesh phantom
Purpose: Accurate dose assessment of patients during a computed tomography(CT) scan is critical in many applications, such as risk projections for optimizationand epidemiological risk assessments. Such CT-scan patient dose assessments
have been conducted by Monte Carlo (MC) simulations coupled withcomputational human phantoms to approximate patient anatomy. Computational
phantoms based on Caucasian data (e.g., ICRP reference phantoms) have beencommonly used but may bias the dose estimates for non-Caucasian patients
because anatomical variation in different races exists. In the present study, we
calculated organ and effective doses using the most advanced mesh-type adultmale and female Korean reference phantoms for several CT scans. These werecompared with those calculated using the ICRP adult mesh-type reference
phantoms.
Methods: To conduct MC simulations, we employed a general-purpose MC
radiation transport code, Geant4. In the Geant4 code, the Korean mesh phantoms
were implemented via G4Tet class, and x-ray beam measurement profiles for the
Siemens 64-slice CT scanner for 120 kVp were simu
lated via
G4VUserPrimiaryGeneratorAcition. Four typical scans (head, chest, abdomenpelvis (AP), and chest-abdomen-pelvis (CAP) scans) were considered to
calculate organ doses and effective doses of the Korean reference phantoms.
The calculation procedure was repeated with the ICRP reference mesh phantoms
for comparison.
Results: Considering the organs outside the scan range, significantly large
differences in the organ doses between the Korean and ICRP phantoms (e.g., bya factor of 9 for the gall bladder at the chest scan) were commonly observed,
but the dose values outside the scan range were very small and thus are not78

#2 Radiation DosimetryO2 - 02

importantly considered. For the organs within the scan range, the organ dose
differences were generally less than 10%, while relatively large differences werealso observed for some cases. For example, the spleen dose of the Korean
male/female phantom was greater by 36% in the AP scan and 23% in the CAP
scan. In the effective dose, the CAP scan showed good agreement between the
Korean and ICRP phantoms with a difference of 5%. Relatively large differenceswere seen in the other scans: 19% in the head scan, 16% in the chest scan, and
18% in the AP scan.
Conclusion: The present study showed that the ICRP reference phantoms could
provide organ doses as well as effective doses different from the Korean
reference phantoms. As the dose assessment directly using the patient CT
images is unavailable, we believe that the use of the Korean reference phantoms
may be more reliable in the dose estimates of Korean patients rather than theconventional approach solely using the Caucasian-based computational
phantoms.
Reference[1] Choi, Chansoo, et al. "Mesh-type reference Korean phantoms (MRKPs) for
adult male and female for use in radiation protection dosimetry." Physics in
Medicine & Biology 64.8 (2019): 085020. [2] Kim, C. H., et al. "ICRP publication
145: adult mesh-type reference computational phantoms." Annals of the ICRP
49.3 (2020): 13-201. [3] Allison, John, et al. "Recent developments in Geant4."
Nuclear instruments and methods in physics research section A: Accelerators,
Spectrometers, Detectors and Associated Equipment 835 (2016): 186-225.

79

Abstracts - Oral

O2 - 03#2 Radiation Dosimetry

Geometric Considerations in L-band EPR Tooth Dosimetry Using
Radiation Treatment Plans

Jong In Park1,*, Jeonghun Oh2, Chang Uk Koo2, In Jung Kim1, Kwon Choi2,Sung-Joon Ye1
Ionizing Metrology Group, Korea Research Institute of Standards and Science,
South Korea1, Department of Applied Bioengineering, Graduate School ofConvergence Science and Technology, Seoul National University, South Korea2
Email: jongin.park@kriss.re.kr
Keywords: EPR, tooth dosimetry, radiation dosimetry
Purpose: In the event of a large-scale radiation exposure, a rapid response tool
is needed for swift dose determination to identify individuals who require
immediate medical attention. This study aims to investigate the feasibility of usinga practical geometric correction (GC) method via mobile phone camera imagesin L-band electron paramagnetic resonance (EPR) tooth dosimetry to improvedose estimation accuracy.
Methods: Beam irradiation was conducted in standard field and radiation
treatment plans for inhomogeous dose irradiation condition. Selected humanincisors were irradiated with calibrated 6 MV photon beams, and verification plans
were generated on CT images of a custom-made poly methyl methacrylate headphantom. The mean doses were calculated using a treatment planning system(TPS) on tooth structures. EPR signal measurements were conducted using ahome-built EPR spectrometer, and the enamel area was calculated from tooth
images acquired using a mobile phone camera. The relative standard
uncertainties in movements of teeth were evaluated using finite element analysis,
and the combined relative standard uncertainties in estimated dose were
estimated with and without GC.

Results: The mean absolute deviations of estimated dose from irradiated dosewere slightly improved from (0.81 ± 0.57) Gy to (0.46 ± 0.33) Gy with GC. The
averaged relative differences between esimated dose and irradiated dose werereduced from 21.0% to 16.8% with statistical significance (p-value = 0.01 < 0.05).
The relative combined standard uncertainties in movements of teeth were 9.1%.

80

#2 Radiation DosimetryO2 - 03

The relative standard uncertainty in estimated dose was evaluated and wellaccordance with RD as a function of irradiated dose.

Conclusion: The use of GC with TPS dose can improve dose estimation accuracy
in L-band EPR tooth dosimetry, making it a potential rapid response tool for swift
dose determination in large-scale radiation exposure events. The study also
highlights the importance of considering uncertainty sources in L-band EPR tooth
dosimetry.
Reference[1] Park J I et al., Dependence of radiation-induced signals on geometry of tooth
enamel using 1.15 GHz electron paramagnetic resonance spectrometer:
improvement of dosimetric accuracy Health. Phys. 2020. [2] ISO. Guide to theexpression of uncertainty in measurement. International Organization forStandardization Geneva, Switzerland, 1995. [3] ISO 13304-1:2020 Radiologicalprotection – minimum criteria for electron paramagnetic resonance (EPR)
spectroscopy for retroscpective dosimetry of ionizing radiation81

Abstracts - Oral

O2 - 04#2 Radiation Dosimetry

GPU-Based Fast Monte Carlo Code for Carbon Ion Therapy

Jin Wu1, Shijun Li1, Yuxin Wang1, Qiang Li3, Xie George Xu2,*
School of Nuclear Science and Technology,University of Science andTechnology of China, Hefei, China 1, Technology Development Department,
Anhui Wisdom Technology Company Limited, Hefei, China 2, School of Nuclear
Science and Technology,University of Science and Technology of China, Hefei,
China 2, Biomedical center,Institute of Modern Physics, Chinese Academy of
Science, Lanzhou, China3
Email: xgxu@ustc.edu.cn
Keywords: GPU Monte Carlo ,Carbon Ion Therapy,Dose calculation,Lineal
energy transfer,TOPAS
Purpose: Although Monte Carlo (MC) simulation methods are the gold standard
for radiation dosimetry, their computational time is often lengthy before achieving
acceptable statistical certainty. This study aims to present a MC dose calculation
code that is both fast and accurate for carbon ion therapy and other irradiation
environments.
Methods: We developed a GPU-based MC package for simulating carbon ion
transport processes in voxelized geometry under the framework of CUDA. The
simulation scheme used Class II condensed history with continuous slowing
down approximation for electromagnetic processes while modeling energy
straggling, multiple scattering and ionization. δ-electrons were terminated with
their energy locally deposited when the primary particle kinetic energy was below
500 MeV/u. Five target nuclei were considered for nuclear interactions, including
Hydrogen, Carbon, Nitrogen, Oxygen, and Calcium. Secondary charged particles
followed the same condensed history scheme. Geant4 was utilized to acquire
total cross section data and secondary particle yield and direction database. Fourkinds of heavy ions (Carbon, Helium, Boron, and Oxygen) were tested in two
different phantoms: a water phantom and a water-bone-water phantom. 1e7
heavy ions were simulated in each test to make sure the mean relative statisticaluncertainty less than 1%. Our code's results were compared with TOPAS.
Results: The results showed great consistency between the two codes. In thewater phantom simulation for Helium, Boron, Carbon and Oxygen ions, therelative errors of precentage depth dose (PDD) from both codes were within 5.0%82

#2 Radiation DosimetryO2 - 04

at all points. In the water-bone-water phantom, the relative errors were all within
5.5%. In addition, we calculated the track averaged lineal energy transfer (LETt)of carbon ion incident water phantom, the relative errors of LETt from both codes
were within 7.0% at all points. Our code's dose calculation time was less than 1
minutes with a single Nvidia GTX 1060 6G card, while TOPAS took about 150
minutes running on the computation platform with the Intel(R) Xeon(R) CPU E54607 with 48 threads.
Conlusion: We developed a GPU-based fast and accurate Monte Carlo dose
calculation code for carbon ion therapy, which also can calculate lineal energy
transfer ( LET) value and be used in dose calculation Helium,Boron and Oxygenion source.

83Abstracts - Oral

O2 - 05#2 Radiation Dosimetry

Preliminary Phantom and Monte Carlo Space Radiation DosimetricStudies for Lunar Missions

Zirui Ye1, Xiaoyu Li1, Zhi Chen1, Jingnan Guo2, Xie George Xu3,*
School of Nuclear Science and Technology, University of Science andTechnology of China, Hefei, China1, School of Earth and Space Sciences,
University of Science and Technology of China, Hefei, China2, School of
Nuclear Science and Technology, University of Science and Technology ofChina, Hefei, China3, Department of Radiation Oncology, The First AffiliatedHospital of USTC, Hefei, China3
Email: xgxu@ustc.edu.cn
Keywords: Space Radiation, Lunar Mission, Radiation Dosimetry, Phantom,
Monte Carlo
In recent years, manned lunar landing projects have been revisited. Humans may
establish bases on the Moon in the future. Space radiation is one of the main
health risks for astronauts, and longer lunar missions will lead to increased risks
for radiation exposure. Due to the complexity of the radiation environments on
the Moon, it is crucial to perform dosimetric assessments using accurate Monte
Carlo methods for appropriate source modeling, human body modeling, andradiation shielding measures. In addition, the ICRP, the NCRP, and NASA haveproposed various methods for calculating dose equivalents for astronautradiation risk assessments. The differences between these methods, which havebeen adopted by many studies, need to be more accurately compared insimulations of lunar missions.
This study aims to compare and refine a Monte Carlo simulation method for
effectively assessing the radiation doses received by astronauts during lunar
surface missions. We employed the TOPAS/Geant4 code to simulate the primaryisotropic incident particles from Galactic Cosmic Rays (GCR) in the upper
hemispherical plane of the Moon surface, as well as albedo particles at different
depths. The particles include electrons, photons, neutrons, pions, and charged
particles with Z ranging from 1 to 28, in the keV to 10 GeV range. The organ
doses and effective dose equivalents deposited within computational humanphantoms after passing through various thicknesses of radiation shielding
devices were calculated, and the flux of secondary particles produced wasanalyzed. The human phantoms used included the first voxelized male phantom
VIP-MAN, ICRP reference phantoms, RPI phantoms of various obesity levels,84

#2 Radiation DosimetryO2 - 05

Chinese-specific adult and minor USTC phantoms, and personalized phantoms
constructed based on medical imaging and deep learning automatic
segmentation. Radiation shielding devices included spherical, hemispherical, andcylindrical models. Dose equivalent calculation methods comprised radiation
weighting factors w_R, different quality factors Q(L) and Q_NASA(Z, E), andRBE-weighted Gray equivalent. In addition to Monte Carlo method results undervarious conditions, the results were also compared with those obtained using flux
to effective dose conversion coefficients.
This study demonstrated the advantages of Monte Carlo simulations, precise
modeling, and appropriate dosimetric assessment methods in accurately
calculating doses in complex radiation environments. Results indicated thatsignificant differences exist between radiation dose results among different
phantoms, and organ doses, excluding skin, varied slightly for the same phantom
under shielding conditions. Although primary heavy ions constituted a small
fraction of the total flux, they contributed significantly to the total dose, while
lunar albedo particles contributing approximately 10% to the total dose.
Compared to the effective dose equivalents calculated using Q_NASA(Z, E), theeffective dose equivalents calculated using radiation weighting factors w_R were
overestimated by about 3 times, while those calculated using Q(L) were
underestimated by more than 15%.
In conclusion, we investigated the impacts of radiation source modeling,shielding modeling, human body modeling, particle transport simulation, anddose equivalent assessment methods on radiation dose and shielding design
accuracy in lunar missions.

85

Abstracts - Oral

O2 - 06#2 Radiation Dosimetry

A script-based automatic commissioning method of virtual
source model for independent dose verification module

Bo Cheng1, Yuan Xu2, Shijun Li1, Qiang Ren3, Xi Pei3, Kuo Men2, Jianrong Dai2,
Xie. George Xu4,*
School of Nuclear Science and Technology, University of Science andTechnology of China, China1, Department of Radiation Oncology, National
Cancer Center/National Clinical Research Center for Cancer/Cancer Hospital,
Chinese Academy of Medical Sciences and Peking Union Medical College,
China2, Anhui Wisdom Technology Company Limited, China.3, School of
Nuclear Science and Technology, University of Science and Technology ofChina, China4, Department of Radiation Oncology, The First Affiliated Hospitalof USTC, University of Science and Technology of China, China4
Email: xgxu@ustc.edu.cn
Keywords: Virtual source model; GPU Monte Carlo; Patient-specific quality
assurance; Independent dose verification
Purpose: A script-based automatic optimization method was developed in thiswork, which can achieve the model commission process automatically within a
few hours.
Methods: The optimization algorithm is based on the three-source virtual source
model (VSM), with a new geometric flatten filter model. Several key parametersof VSM are used as iterative variables, which include maximum photon energy
E_max, the average electron energy E_e, the weight of each source weight, thecorrection factor h for different flatten filter, and the off-axis softening factor
δb_pri of primary photons. The objective function δ_PDD is used to represent thedose difference in PDD curve, δ_OF to represent the dose magnitude difference(output factor), and δ_LAT to represent the dose difference in lateral curve. Theenergy parameters E_max and E_e were used to control the longitudinal PDD
curve because higher energy particles deposit higher dose at deep depth andlower dose at shallow depth. δ_PDD is the objective function of E_max, andbecause electrons are mainly deposited at shallow depths, the objective functionof E_e is the δ_PDD at shallow depth. The weight of primary and secondary
photon sources were used to control the relative dose magnitude (output factor)
in different fields. The off-axis softening coefficient δb_pri represents therelationship between the particle energy distribution and the off-axis distance,
which was used to control the lateral dose profile. A correction factor h is used86

#2 Radiation DosimetryO2 - 06

to represent the geometric differences of flatten filter. The objective function ofδb_pri and h is δ_LAT to control the lateral dose profile.
Then an automatic optimization workflow is performed for each key parameter,
including commission data pre-processing, automatic dose calculation in watertank, automatic plotting, and automatic updating of model parameters. The
parameter optimization order is E_max, E_e, weight, δb_pri, and h, forming a loop.
Four times optimization needs to be performed for each parameter, and the
optimal solution is selected as the base solution for the next parameter
optimization.
Results: More than ten accelerator models have been built including the Varianand Elekta accelerators, with the Pinnacle or Monaco treatment planning system.
For an input commission data, the optimization can be done automatically within
7 hours without any human intervention. The relative dose difference of all models
was within 2% compared to the measured tank data. ArcCHECK measurements
and ArcherQA calculated dose in homogeneous phantoms were compared forclinical plans, and the average gamma passing rates (3%/2mm) were higher than
95% in all accelerators. With the GPU accelerated, dose verification can becompleted in two minutes for each clinical plan.
Conclusion: The proposed optimization method can achieve an automaticcommissioning process in a few hours, which can generate accurate, directly
usable accelerator models for clinical dose verification.
Acknowledgement
This work was supported, in part, by the following grants: University of Scienceand Technology of China (USTC) (#YD2140002002 and #XDB39040600), Natural
Science Foundation of Anhui Province, China (#1908085MA27), National Natural
Science Foundation of China (#12275372), and Beijing Hope Run Special Fundof Cancer Foundation of China (#LC2022B25).

87

Abstracts - Oral

O2 - 07#2 Radiation DosimetryAge-dependent Iodine-131 S values Based on ICRP Reference
Pediatric Phantoms

Soo Min Lee1, Yumi Lee1, Ji Won Choi1, Chul Hee Min1, Hyun Joon Choi2,
Chansoo Choi3, Yeon Soo Yeom1,*
Department of Radiation Convergence Engineering, Yonsei University / Korea1,
Department of Radiation Oncology, Yonsei University Wonju College of Medicine / Korea2, J Crayton Pruitt Family Department of Biomedical Engineering,
University of Florida, FL / USA3
Email: ysyeom@yonsei.ac.kr
Keywords: S values, I-131, Computational phantom, thyroid, Monte Carlo simulation, pediatric voxel phantoms
Although radioactive iodine therapy is beneficial to deliver the prescribed dose to
the thyroid in patients with hyperthyroidism or thyroid cancer, it is accompaniedby the unwanted dose delivery to other normal organs and tissues, potentiallyleading to undesired late health effects such as radiation-induced cancer. The
prediction of late effects should be preceded by estimating the organ/tissue absorbed doses of the patients, but the patient-specific anatomy information for alarge-scale patient cohort is usually unavailable. Alternatively, the iodine-131 S
values (rT ← thyroid) estimated using the reference adult phantoms of the International Commission on Radiological Protection (ICRP) have been generally used
for adult patients, but may not be applicable for pediatric patients. In the present
study, therefore, we produced a new dataset of age-dependent iodine-131 S
values using the ICRP reference pediatric phantoms recently released in ICRP
Publication 143.
The age-dependent iodine-131 S values were calculated for the 30 target organs
and tissues which are considered radiosensitive by the ICRP. First, the pediatricphantoms were implemented into the Geant4 Monte Carlo code using the
G4VNestedParameterization class. We simulated gammas and betas of the iodine-131, the energies and yields of which were extracted from the decay dataof ICRP Publication 107, to be uniformly emitted over the thyroid using theG4UserPrimaryGeneratorAction class. The number of primary particles varied
from 109 to 1010 depending on the age to keep the relative errors of the calculated values below 5%. A secondary production range cut value of 1 mm wasapplied via the G4EmLivermorePhysics library. Finally, the energies deposited in88

#2 Radiation DosimetryO2 - 07

the target organs and tissues were calculated using the G4PSEnergyDeposit
class.
The produced age-dependent iodine-131 S values for most target organs andtissues tended to increase as the age decreases. For example, the S value forthe stomach at newborn was 1.47, 3.55, 6.67, and 11.13 times greater than thevalues at 1 year, 5, 10, and 15 years. This trend can be simply understood considering the younger phantom generally shows shorter inter-organ/tissue distances. We also found that the pediatric S values were generally greater thanadult values from the adult phantoms. A few organs and tissues near the thyroid,
however, exhibited an unexpected tendency, due mainly to the anatomical variation of individuals inherent in the ICRP reference phantoms. For example, the S
value for the oral mucosa showed the similar values at 1, 5, and 15 years with adifference of less than 3%. The S value at 15 years was even 1.27 times greaterthan that at 10 years, which was the lowest.
The age-dependent iodine-131 S values produced in the present study may
contribute to the significant improvement of the organ/tissue dose estimates ofpediatric patients treated by radioactive iodine therapy, eventually supporting theclinical decision-making and epidemiological studies for the medical staffs andresearchers.
Acknowledgement
This work was supported by the Basic Science Research Program through the
National Research Foundation of Korea (NRF) funded by the Ministry of Education(NRF-2020R1C1C1008093).

89

Abstracts - Oral

O2 - 08#2 Radiation Dosimetry

CPU-GPU Coupling Monte Carlo Acceleration Method for Dose
Calculation of Mesh-type Phantom
Shuchang Yan1, Rui Qiu1,*, Xiyu Luo1, zhen Wu1, Hui Zhang1, Junli Li1
Department of Engineering Physics, Tsinghua University/China1
Email: qiurui@tsinghua.edu.cn
Keywords: Fast Monte Carlo, Mesh-type Phantom, Heterogeneous
Currently, mesh-type phantom dose calculation requires conversion into tetrahedral mesh for dose calculation in advance. Although its calculation speed iscomparable to that of existing voxel geometry, it is far from meeting the requirements of real-time calculation. While mainstream GPU parallel computations
have been carried out for voxel geometry, research in the field of mesh-typemodel geometry is still lacking.
The development of a Monte Carlo method and parallel algorithm that couples
CPU and GPU to accelerate mesh-type phantom dose calculations is of greatsignificance. Based on particle transport algorithms and underlying code of serial
Monte Carlo programs, this paper presents a CPU-GPU coupling Monte Carlo
program, which can perform dose calculations not only for tetrahedral phantom
converted from polygonal surface phantom but also for mesh-type phantom directly.
With the advantages of GPU in geometric processing and incorporating opticaltracking technology and efficient data structure, the efficient geometric positioning and intersection calculation algorithms have been developed, enabling particle transport in polygonal and tetrahedral meshes on GPU device. To further
speed up the calculation, the program is optimized through several techniques,
including (1) reasonable allocation of computing resources to ensure effective
usages of both CPU and GPU, (2) cross-boundary algorithms to manage theparticle transport between identical materials, and (3) efficient data storage tominimize the time consumption of data access.
Doses of the adult mesh-type reference computational phantoms (MRCPs) under
different energy and standard external irradiation conditions are calculated bythis program and Geant4 respectively, and results show good agreement with
Geant4 with relative deviation within 5%. Compared to Geant4, the program
achieves a significant increase of about 30 times in calculation speed. It canaccelerate dose calculations for both mesh-type phantom and tetrahedral phantom. This approach has great potential to further improve the efficiency of particle
transport simulations and achieve real-time dose calculations of mesh phantom
with integration of the technique of phantom deformability.

90

#2 Radiation DosimetryO2 - 09#2 Radiation Dosimetry

Personalized Cumulated Activity Distribution Calculation using Biokinetic Compartmental Model in PET Imaging

He Yutao1, Chen Zhi1,*
School of Nuclear Science and Technology, University of Science andTechnology of China, China1
Email: zchen@ustc.edu.cn
Keywords: Cumulated activity, Biokinetic, Internal dosimetry, PET/CT
Purpose: Accurately calculating the cumulated activity distribution of radiopharmaceuticals in the human body is essential for assessing the internal radiationdose in PET imaging. However, in most PET imaging studies, only PET images
at one single time point are available, without kinetic information of the biodistribution of radiopharmaceuticals. In terms of this problem, many methods weredeveloped to estimate cumulated activity distribution, and the main methods are
as follows. The first one assumes that the radiopharmaceuticals are immediately
taken up after injection without considering their transfer in the body, and only
considers the physical half-life, referred to as ODC (Only Decay Concerned
Method). The second one uses experimental statistical data on the distributionof radiopharmaceuticals in various organs to construct cumulated activity distribution, referred to as ESD (Experimental Statistical Data Method). And the last
one calculates cumulated activity based on a biokinetic compartmental model,
referred to as BCM (Biokinetic Compartmental Model Method). The aim of thisstudy is to compare the three methods, and then calculate the personalized cumulated activity distribution of 18F-FDG patients by using BCM based on a single
time-point PET image.
Methods: The calculation of cumulated activity distribution by BCM involves thefollowing steps: Firstly, a compartmental model for the organs of interest is built,
the variation of activities in each compartmental is described as a set of differential equations (ODE); Then, the PET image data at a specific time point of eachpatient is used as a constraint to fit and optimize the biokinetic parameters of thecompartmental model by using the simulated annealing algorithm. The solved
ODE system provides a time-activity curve for each compartment in the biodistribution model. Finally, the cumulated activity of each voxel corresponding toeach organ is calculated based on the patient's PET image activity distribution
information. The data from 15 cases were analyzed and calculated, and the91Abstracts - Oral

O2 - 09

model calculation results were compared with the dynamic PET experimental dataof Hu et al. (2021).
Results: Comparing the results with ICRP 128 data, ODC significantly underestimated the cumulated activity of organs with a frequent material exchange, suchas heart and bladder, with a relative deviation range of about -90% to -10%. As
for ESD, using the residence time data from Hu et al. (2021) as a reference, therelative deviation of the total cumulated activity was only 2.45%. However, themaximum relative deviation for major organs such as the lungs and brain was
86.19%, with an average relative deviation of about 37%. For the 15 cases analyzed in this study, the calculation results of BCM generally showed a good
agreement with experimental data, with an average relative deviation within ±15%.
However, some patients' results perform higher deviations than the average value,and further analysis of the uncertainty of the biodistribution model results is necessary.
Conclusion: A biokinetic compartmental model is used to calculate personalized
cumulated activity distribution for specific patients. However, some patients' results perform higher deviations, and further analysis of the uncertainty of thebiodistribution model results is necessary.

92

#2 Radiation DosimetryO2 - 10#2 Radiation Dosimetry

Analysis of Occupational doses in Korean nuclear power plants
over the past ten years

Chang Ju Song1, Tea Young Kong1,*
School of Chosun University/Korea1
Email: tykong@chosun.ac.kr
Keywords: Occupational doses, Korean nuclear power plants, Dose constraints
The International Commission on Radiological Protection (ICRP) issued ICRP 103
recommendation in 2007, recommending the application of dose constraints tooptimize radiation protection to resolve exposure inequity among radiation workers. This study analyzed the annual occupational doses in Korean nuclear power
plants (NPPs) over the past ten years to find their dose distribution. As a result,it was found that NPP workers received annual average occupational doses of
0.3-0.8 mSv. These doses are much lower than the annual occupational dose
limit of 50 mSv stipulated in the Nuclear Safety Act. In addition, most of the NPP
workers received less than 0.1 mSv/y, but some NPP workers received relatively
higher doses than others. Korean regulatory body conducted a study on the application of dose constraints to resolve exposure inequity and determine dose
constraints (draft) for each radiological occupation.
This study analyzed the occupational doses in Korean NPPs for the last ten years(2012-2021) to be used for optimizing the occupational doses in the future. Occupational doses in Korean NPPs could be obtained from the annual reports for
NPP radiation control. These annual reports enabled the analysis of internal andexternal exposure doses and occupational doses depending on the NPP operating conditions, including normal operation, planned maintenance, and intermediate maintenance.
As a result of analyzing the ratios of internal and external exposure doses, it was
found that external and internal exposure in pressurized water reactors accounted
for 95-98% and 2-5%, respectively. For pressurized heavy water reactors, theexternal and internal exposure accounted for 76-84% and 14-26%, respectively.
Both results showed that external exposure was much higher than internal exposure in Korean NPPs. Furthermore, the analysis results of occupational doses
depending on the NPP operating conditions revealed that most radiation exposures occurred during planned maintenance periods, approximately 85-91%.
Radiation exposures during normal operation and intermediate maintenance periods accounted for 9-14% and 0-4%, respectively. Therefore, the results of thisstudy provide occupational doses in Korean NPPs and can be used in the future
to optimize the doses of workers who receive relatively higher doses than others.
93Abstracts - Oral

O2 - 11#2 Radiation Dosimetry

Development of Virtual Dose Monitoring System for Interventional
Radiology (VIR)

Gahee Son1, Haegin Han1, Sungho Moon1, Bangho Shin1, Hyeonil Kim1,
Suhyeon Kim1, Soon Young Song2, Chan Hyeong Kim1,*
Department of Nuclear Engineering, College of Engineering, Hanyang University/ Korea1, Department of Radiology, Hanyang University Hospital / Korea2
Email: chkim@hanyang.ac.kr
Keywords: Interventional radiology, occupational dose, computational humanphantomInterventional radiologists are routinely exposed to stray radiations during theprocedures and thus receive relatively high doses compared to other medical
staffs, especially to eye lens and skin which are not shielded by the lead apron.
The physical individual dosimeter worn under the lead apron does not properly
measure the dose to the eye lens and the skin exposed. To address this issue,in the present study, a virtual dose monitoring system for interventional radiology(VIR) was developed, which can estimate effective dose and organ doses in realtime reflecting continuously moving radiologists and radiation field in the operation room. The VIR system consists of three modules: (1) machine tracking module, which tracks the C-arm and lead glass shield; (2) body tracking module,
which tracks the motion of the radiologists and continuously generates the computational human phantoms of the radiologists; and (3) dose calculation module,
subdivided into Real-time Dose Calculation (RDC) and Post-procedure Dose
Calculation (PDC) submodules, which calculated the doses to the radiologists.
In the present study, the developed VIR system was tested for a simulated operation conducted in an operating room using a C-arm fluoroscopy machine (Allura
Xper FD20, Philips, Netherlands). The test showed the capability of the machine
and body tracking modules to acquire the required data in real time, even thoughbody tracking did not work very well when there are multiple radiologists working
together. The dose calculation module operated as intended; that is, the RDC
submodule displayed the skin dose distribution and eye lens dose in real time,
and the PDC submodule properly calculated effective dose and organ doses bytransporting particles in Geant4. The intercomparison between the two submodules generally showed good agreement in the skin and eye lens dose values.

94

#2 Radiation DosimetryO2 - 12#2 Radiation Dosimetry

Dose coefficients for soil contaminations produced using meshtype Korean reference phantoms (MRKPs)Ji Won Choi1, Yumi Lee1, Bangho Shin2, Chansoo Choi3, Yeon Soo Yeom1,*
Department of Radiation Convergence Engineering, Yonsei University/ Korea 1,
Department of Nuclear Engineering, Hanyang University/ Korea2, J Crayton
Pruitt Family Department of Biomedical Engineering, University of Florida/
U.S.A.3
Email: ysyeom@yonsei.ac.kr
Keywords: Mesh-type Korean Reference Phantoms, soil contamination, dosecoefficients, Monte Carlo[Purpose]
The International Commission on Radiological Protection (ICRP) Publication 144(2020) recently released a comprehensive dataset of dose coefficients (DCs) for
external exposures to environmental sources (soil, air, and water) using the ICRP
reference phantoms based on Caucasian data. The use of the ICRP-144 DCs toestimate organ doses or effective doses of non-Caucasian populations may bequestionable considering the fact that the anatomical difference in different races
exists. In the present study, Korean-specific DCs for soil contamination were
produced using the adult male and female Mesh-type Korean Reference Phantoms (MRKPs) and then compared with the ICRP-144 DCs based on the Caucasian data.
[Methods]
Using a general-purpose Monte Carlo (MC) radiation transport code, Geant4, the
MRKPs were irradiated by simulating the photon and electron soil sources basedon the phase-space data, which were used in the calculation of the ICRP-144
DCs. The number of primary particles transported for each simulation varies from108 to 1010 to keep the statistical uncertainties of the calculated DCs for organ
doses below 10% and those for effective doses below 1%.
[Results]
For photons, the organ DCs of the MRKPs tend to show a good agreement with
the ICRP-144 DCs (deviations < 20%) for most energies. At energies below 0.05
MeV, however, significant differences were observed by up to a factor of 17.4
(for gall bladder at 0.02 MeV). The Korean DCs for most organs showed smaller95

Abstracts - Oral

O2 - 12

values than the ICRP-144 DCs except for several organs (e.g., lungs, thyroids,
and muscle). For electrons, notable differences in the organ DCs between Korean
and Caucasian were generally observed over the entire energy region (deviations> 20% for most cases). The effective DCs of the MRKPs for photons showed an
excellent agreement with the ICRP-144 DCs; the deviations were all less than
14%. For electrons, the Korean effective DCs again showed notable differences
by up to 78% especially at the energy region between 0.06 and 1.0 MeV.
[Conclusion]
To the best of our knowledge, we first produced a comprehensive dataset of the
Korean-specific DCs for photon and electron soil contamination sources using
the most advanced mesh-type Korean reference phantoms. We believe that the
Korean DCs together with the ICRP-144 DCs may help not only to estimate morereliable effective doses dedicated to the Korean population for radiation protection purposes, but also to improve the Korean individual organ dose assessment
required for prospective and/or retrospective risk assessment.

96

#2 Radiation DosimetryO2 - 13#2 Radiation Dosimetry

Patient-Specific Dose Assessment for Nuclear Medicine Therapy
Based on 177Lu-PSMA

Ziyi Hu1, Ruiyao Ma1, Rui Qiu1,*, Ankang Hu1, Zhen Wu2, Hui Zhang1, Junli Li1
Department of Engineering Physics, Tsinghua University, Beijing, China1,
Nuctech Company Limited, Beijing, China2
Email: qiurui@mail.tsinghua.edu.cn
Keywords: Nuclear medicine therapy, Patient-specific dosimetry, Monte Carlo
Purpose: Nuclear medicine therapy is an important means of using radionuclides
to treat cancer and other diseases, especially for middle and advanced tumors,and has important applications in the treatment of bone metastases, thyroid
cancer, and other diseases. Patient-specific dose assessment for clinicalpatients is the basis for specific treatment planning that meets precise medicine
requirements. Based on clinical patient data of 177Lu-PSMA, this study
established a personalized pharmacokinetic model in patients, and conducted astudy on the accurate calculation of dose distribution in patients.
Methods: This study proposed a new method for patient-specific dose
assessment of nuclear medicine patients based on nuclear medicine images:
SPECT/CT data of clinical nuclear medicine patients were collected at first. The
spatial distribution of drugs in the above functional images was registered withthe anatomical images of patients after organ segmentation to obtain thechanges of drug activity in patients' organs over time, and then the individualized
pharmacokinetic model of patients was established based on the compartment
model. The individual phantom of one of the patients was established based on
the anatomical image, and the individual dose was evaluated from the organ
scale and voxel scale respectively. And a neural network model suitable forindividual dose evaluation was established by using the neural network method.
Results: The maximum difference in the mass of radiosensitive organs between
the patient-specific phantom and the reference phantom is more than 200%, andthe location and shape of some organs are also different. The dose calculation
results show that for 177Lu internal irradiation of this patient, the organ dose
difference between the patient-specific phantom and the reference phantom is
up to 40%, and the difference of effective dose is 28%.
Conclusion: This study will provide a tool for patient-specific dose assessmentof nuclear medicine patients, and provide an important reference for clinicians tomake diagnosis and treatment plans for individual patients.

97

Abstracts - Oral

O2 - 14#2 Radiation Dosimetry

Establishment and first application of 4D dose calculation in
China

Zhaoxing Liu1, Ri Zhao2,*, Na Liu1, Jing Zhang2, Runcheng Liang2, Xin Liu1
China Institute for Radiation Protection, Taiyuan, China1, Shanxi Provincial Key
Laboratory for Translational Nuclear Medicine and Precision Protection,
Taiyuan, China2
Email: zhaor.abc@163.com
Keywords: 4D dose, Mesh phantom, Motion capture, Tetrahedralization,
Accurate protection
Radiation dose calculation is a key technology for prospective and retrospective
occupational dose monitoring which is mainly realized based on the human digital
phantom and Monte Carlo(MC) simulation method. At present, voxel phantom ismost used which cannot be posture-deformed, thus the dose results are in 3D
and static unable to meet the needs of precise protection in future nuclear facility
operation, nuclear technology, and medical application. Recently, the emergenceof a novel mesh-type phantom, motion capture technique, and high-speed MC
simulation method make it possible for 4D dose calculation.
Here, we have established the 4D dose calculation process in two steps: firstly,
posture deformation of MRCP (adult mesh-type reference computational
phantom proposed by ICRP) is carried out through rigid rotation, VGL (Volume
Graph Laplacian) operator and ARAP(As Rigid As Possible) algorithm; secondly,
based on Delaunay algorithm, the deformed mesh-type phantom is
tetrahedralized and imported into Geant4 code. Moreover, a comprehensive 4D
dose calculation system is developed by integrating the Kinect motion capturecommercial module and a GPU-based cloud computing platform.
A verification experiment was carried out using the combination of the Female
Rando physical phantom and a BOMAB phantom to be irradiated in a Co-60
standard radiation field. By combining the two phantoms, two postures, i.e. astanding posture with hands crossed and a squatting posture, were constructed.
TLD elements were placed at the central spots of the heart, lung, liver, and ovary
inside the phantom, as well as Hp(3) TLD dosimeter beside the eye and Hp(10)TLD dosimeter on the chest. The deviation between the measured and calculated
doses were14.8% (kidney), 9.7%( Hp(3)), 3.3%( Hp(10)),for the standing posturewith hands crossed.,and 18.8% (liver) 9.3%( Hp(3)) and 10.1%( Hp(10)) for thesquatting posture98

#2 Radiation DosimetryO2 - 14

The first application of the proposed technique to a nuclear facility in China was
implemented at Qinshan NPP. The motion of a worker was captured when he wassetting up radiation shielding, and the radiation source was characterized by agamma-ray spectrometer. The comparison between the measured (by TLD) andcalculated doses showed that the deviation was less than 10% for Hp (10), andless than 15% both for Hp (3) and Hp (0.07). The result confirmed the feasibility
and reliability of in-situ 4D dose calculation.

99

Abstracts - Oral

O2 - 15#2 Radiation Dosimetry

Establishment of High Dose Rate Pulsed γ Radiation Facility

Zhihui Tang
China030006,China Institute for Radiation Protection1
Email:
Keywords: Pulsed Gamma Radiation, Steady-state Radiation Field Cutting,Pulsed
Gamma Measurement
There are a large number of critical safety alarms in nuclear fuel plants and reprocessing plants. Because these critical safety alarms lack relevant radiation
performance testing devices, it is urgent to establish pulse radiation standard
devices to solve the problem of measuring and transmitting critical safety alarms.
According to IEC60860:2014, the average air specific release kinetic energy rateof the minimum criticality accident is 12Gy/h, and the pulse width of pulse radiation used for testing critical safety alarming devices is 1 ms～3 s, so the pulse
gamma radiation device should be able to generate pulse gamma radiation witha pulse width of millisecond and an average air specific release kinetic energy
rate of Gy/h.
The foundation of the pulsed γ-ray irradiation device is the standard γ-ray device
with therapeutic level, and the radioactive source is 5000Ci 60Co. After analyzing
the existing pulse γ radiation devices, the design scheme of using the beam
limiting diaphragm slider to drop down to generate pulse γ radiation is determined.
The thickness of the slider in the design scheme is preliminarily determined by
Monka simulation, and it is also proved that the design scheme can generate
single pulse γ radiation similar to square wave. On the basis of the design scheme,
the pulsed γ radiation device is established. The preliminary test and debugging
results show that the shielding effect of the pulsed γ radiation device meets therequirements, and the pulsed γ radiation device can operate normally. On thebasis of the normal operation of the pulsed γ radiation device, the specific kinetic
energy of air, the uniformity of pulsed γ radiation field and the pulse width ofpulsed γ radiation were measured. The measurement method of air specific release kinetic energy is to measure the accumulated charge of pulsed γ-radiationand then convert the accumulated charge into air specific release kinetic energy.
Before measuring the air specific release kinetic energy of pulsed γ radiation,pulse X-ray is used to verify the time resolution of ionization chamber. The measurement method of field uniformity is to measure the change of accumulated
charge by vertically and horizontally moving ionization chamber. The pulse width100

#2 Radiation DosimetryO2 - 15

measurement method is to use an infrared timer to measure the time when thepulse interval of the slider passes the reference point of the main support when
the slider falls.
The measurement results show that: 1) The air specific release kinetic energy ofsingle pulse γ radiation is between 136 μGy and 311 μGy, and the repeatabilityof the measurement results is less than 1.3%.2)The horizontal radiation field diameter of pulsed γ radiation field is 10cm, and the vertical radiation field diameteris 9.5cm. The horizontal radiation field is symmetrically distributed, and the vertical radiation field radius is large and small, showing asymmetry. The reason may
be the continuous acceleration of the slider. It is known that the pulse width rangeof pulse gamma radiation is 80ms～220ms, and the average air specific releasekinetic energy rate of pulse gamma radiation is 5.1Gy/h～7.1Gy/h combined withthe measurement data of air specific release kinetic energy.

101Abstracts - Oral

O2 - 16#2 Radiation Dosimetry

Establishment of radiation biological dose estimation curve basedon automatic analysis of dicentric chromosomes

Li Shufang1, Qin Xiujun2, Yin Jingjing3, Xu Wenli4, Zhang Yajuan5, Yue Juan6, Li
Jianguo7,*
Department of Radiology and Environmental Medicine, China Institute for
Radiation Protection1
Email: ljg2547@163.com
Keywords: curve; radiation biological dose estimation; automatic analysis ;dicentric chromosomes.
Purpose: Objective To establish the radiation biological dose estimation curve
based on automatic analysis of dicentric chromosomes made by Hui Ronghe
Science and Technology Ltd.in BeiJing.
Methods ::Blood samples were collected from 5 healthy people, and were
irradiated with gamma ray at the dose of 0, 1, 2, 3, 4, 5Gy and the dose rate was
17.4Gy/h. After 48 hours of whole blood culture, the samples were harvested andchromosome slides were prepared. Chromosome manual analysis, Huirong andautomatic analyzer analysis, manual calibration after instrument analysis(excluding false positive of double centromere) were carried out for themetaphase division phase of the specimen, and the results were fitted and theestimation curve was established. Finally, three curves were used to estimate theblind dose.
Results:The curve equation established by manual analysis is y=3.132X22.7015x+1.3651, R2=0.9962. The curve equation established by automatic
instrument analysis is y=1.4192X2+2.8576x+10.665, R2=0.9992. The equationof the curve after manual calibration is y=0.9858X2+1.5074x+0.5646,
R2=1; Blind sample estimation deviation is the smallest curve established byautomatic instrument analysis, all of which are less than 10%.
Conclusion:The deviation between the curve calculation results of the analyzer
and the actual situation is acceptable. The instrument and the corresponding
curve can be used to estimate the dose of 1-5Gy irradiated samples.

102

#2 Radiation DosimetryO2 - 17#2 Radiation Dosimetry

A performance study of a new personal neutron dosimeter basedon SSNTD

Yan Jiao1, Xiaofeng Wei1, Qinjian Cao1,*
China Institute for Radiation Protection/China1
Email: caoqinjian@163.com
Keywords: external exposure，neutron，personal dosimeter，CR-39
As a passive neutron personal dosimeter, CR-39 solid state nuclear track
detector is convenient, cheap, insensitive to gamma and beta rays and hasstrong anti-interference ability and better response to fast neutrons.A newly
neutron personal dosimeter has been developed combining with two
measurement methods of CR-39 and albedo thermoluminescence.Performanceexperiments are carried out on the CR-39 detector according to the international
standard ISO21909-1. The results show that the CR-39 solid state nuclear track
detector has good repeatability, batch homogeneity, dose linearity, stability,
response of reference radiated field and no photon-sensitivity.The detection
threshold and uncertainty of dose measurement are presented below.It provides
an experimental data reference for the introduction of international standard.

103

Abstracts - Oral

O2 - 18#2 Radiation Dosimetry

Fast and Accurate GPU-Based Monte Carol Independent Dose
Verification Platform for Proton Therapy

Shijun Li1, Li Tao1, Xi Pei2, XieGeorge Xu3,*
School of Nuclear Science and Technology, University of Science andTechnology of China, Hefei, China1, Anhui Wisdom Technologies, Inc., Hefei,
China2, School of Nuclear Science and Technology, University of Science andTechnology of China, Hefei, China3, Department of Radiation Oncology, The
First Affiliated Hospital of University of Science and Technology of China,
Hefei, China3
Email: xgxu@ustc.edu.cn
Keywords: monte carlo, GPU-based, quality assurance, proton therapy
Purpose: A GPU-based Monte Carlo (MC) independent dose verification platform
for pencil beam scanning (PBS) proton therapy was proposed, and integrated
into the commercial software ArcherQA to achieve fast and accurate quality
assurance.
Methods: Proton transport was modeled by a Class II condensed history algorithmwith a continuous slowing down approximation. Ionization, multiple scattering,energy straggling, elastic and inelastic nuclear interactions were included. The
well-known accurate but time-consuming MC code TOPAS was applied tovalidate our transport model implementation. The beam model was establishedby matching our code simulation with measured data, including in-air lateral dose
profiles, integrated depth dose, and the absolute dose of 19 beam energies. Our
re-calculated 3D dose were based on patient-specific treatment plan exported
from treatment planning system (TPS). Our code was implemented under thecompute unified device architecture (CUDA) platform to take full advantage ofGPU extensive thread-level parallelism and high efficiency, and integrated into
the commercial software ArcherQA to realize automatic dose re-calculation forclinical use, including data preparation, dose calculation, 3D gamma analysis.
Results: By using dose distributions produced with TOPAS to validate our resultsin various material phantoms, good agreements were found. The 3D-gamma
pass rate with a 2 mm/2% criterion in the region with dose greater than 10%
maximum dose were computed to be over 99.2%, which clearly demonstrated
the accuracy of our transport model. Between our simulation and measurement,
the distal range differences at 80% (R80) and 20% (R20) dose levels for eachenergy were both below 0.05 mm, and point-to-point relative error of integrated

104

#2 Radiation DosimetryO2 - 18

depth dose in water tank were below 1%. 40 patient cases ranging from foot tohead were study. The mean gamma pass rate were 99.5 between our code and
TOPAS, 98.2 between our code and TPS, and 97.8 between TPS and TOPAS.
With a single NVIDA TITAN V card, the computational time is 0.82-4.54 s for
1×107 proton histories, depending on phantom and energy. Compared to tensof CPU hours for TOPAS, the speed improvement is extremely large.
Conclusion: The proton beam model built in our code is accurate and trustworthy.
ArcherQA can perform independent dose verification with high accuracy andefficiency for proton therapy.

105

Abstracts - Oral

O2 - 19#2 Radiation Dosimetry

Demonstration of Realtime Monte Carlo Person-specific Organ
Dose Computing involving Automatic Segmentation of 10,000
Patients

Zirui Ye1, Yang Lu2, Haoran Zheng3, Bei Yao3, Xi Pei2, Xie George Xu4,*
School of Nuclear Science and Technology, University of Science andTechnology of China, Hefei, China1, Anhui Wisdom Technologies, Inc. Hefei,
China2, Health Center, The First Affiliated Hospital of USTC, Hefei, China3,
School of Nuclear Science and Technology, University of Science andTechnology of China, Hefei, China4, Department of Radiation Oncology, The
First Affiliated Hospital of USTC, Hefei, China4
Email: xgxu@ustc.edu.cn
Keywords: Computed Tomography, Organ Dose, Person-specific, Realtime
Monte Carlo, Automatic Segmentation
Computed Tomography (CT) has become a primary medical imaging modality for
disease screening and diagnosis, experiencing exponential growth in hospitalsettings. However, CT scans yield radiation doses approximately 10 times higher
than conventional X-ray imaging, resulting in increased collective radiation
burden and cancer risk for the public. Consequently, there is an urgent need for
dose management and tracking analysis for patients undergoing CT examinations.
Recent advances in Artificial Intelligence (AI) and GPU parallel computing haveshown potential in the field of dosimetry, offering unprecedented opportunities
to transition from the traditional ICRP population-average reference phantombased dosimetry system to a next-generation "personalized" dosimetry system.
This study aims to develop a standardized method for rapid and precise
estimation of "personalized" organ radiation doses during CT scans in hospitals
and to create a software system for convenient hospital adoption and promotion.
We utilized a convolutional neural network (CNN)-based CT image automaticorgan segmentation software called DeepViewer to construct personalized
phantoms for patients, and employed GPU-accelerated fast Monte Carlo dose
calculation software, ARCHER-CT, to accurately simulate X-ray transport anddosimetry for validated CT scanner models. The radiation doses received by thepatients' primary organs were analyzed. A preliminary user-end CT dose
calculation and analysis software platform was also developed based on the106

#2 Radiation DosimetryO2 - 19

dosimetry engine, and it was tested on 10,000 patient data from the Health Centerof the First Affiliated Hospital of the University of Science and Technology ofChina.
Our CT organ dose calculations on the patients' DICOM data demonstrated that,compared to the current ICRP reference phantom-based methods and"individualized" phantom dosimetry calculations tailored to different body sizes,
our "personalized" phantom approach, incorporating patient-specific anatomical
information, improved organ dose calculation accuracy by approximately 40%.
The automatic segmentation for each personalized phantom required about 60
seconds, while ARCHER-CT Monte Carlo computation took approximately 30
seconds. The analysis of 10,000 CT patient cases demonstrated the feasibility
and advantages of our "personalized" dosimetry approach in dose accuracy.
This study addresses three key technical challenges to achieve "personalized"organ radiation dose calculation for CT patients: 1) rapid construction of
personalized computational phantoms, 2) accurate and fast radiation-sensitive
organ and tissue dose calculations, and 3) quick and automated collection andprocessing of CT scan information in big data scenarios.

107Abstracts - Oral

O3 - 01#3 Radiation Detection and Sensor Technology

Development of Two Types of Lightweight Portable Neutron Survey Meter

Takashi Nakamura1, Tomoya Nunomiya1,*, Seini Yamamura1
Fuji Electric Co. Ltd.1, Tohoku University2
Email: nunomiya-tomoya@fujielectric.com
Keywords: Lightweight, Neutron, Surveymeter
We have developed two types of lightweight active portable neutron survey metercomprising a proportional gas counter with a mixture of methane and nitrogenfor measuring the ambient neutron dose equivalent H*(10). One type without using polyethylene moderator has already been published in Journal of Nuclear
Science and Technology. Since no heavy polyethylene moderator is used, thisneutron survey meter is only about 2.2 kg in weight causing a weight reduction
over 70 % compared with the conventional moderated-type survey meter. The
neutron energy response to H*(10) showed very good agreement within about
30% to quasi-mono-energetic and continuous energy neutrons in actual workplace neutron fields over wide energy region from thermal energy up to about
200 MeV. This survey meter can be used in various accelerator facilities. The
neutron energy response of this survey meter was however highly deviated fromH*(10) to the mono-energetic neutron fields in the keV region. Then, we havedeveloped another type of lightweight neutron survey meter with inserting a thin
(2.5 cm thick) polyethylene moderator in the previous gas counter. This new survey meter has a little heavier weight of 3.5 kg, but still lighter than the conventional polyethylene moderated survey meters of about 7 to 11 kg. This survey
meter highly improved the neutron energy response in the keV region to the fluence–to-dose equivalent conversion coefficient and can be appropriately usedin nuclear power plants where low energy neutrons occupy mostly. These two
types of lightweight neutron survey meters are now commercially available andare widely used in various nuclear facilities.

108

O3 - 02#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

Detection of gamma-ray photons with halide lead perovskite single crystals for high energy resolution

Xin Wang1,*, Yubing Xu1, Yuzhu Pan1, Omolola Esther Fayemi, Qing Li, Wei Lei
School of Electronic Science and Engineering, Joint International ResearchLaboratory of Information Display and Visualization, Southeast University,
Nanjing, China1, Department of Chemistry, School of Mathematics and PhysicalSciences Faculty of Natural and Agricultural Sciences North-West University,
Mafikeng Campus, Private Bag X2046, Mmabatho 2735, South Africa2
Email: xinw@seu.edu.cn
Keywords: Perovskite single crystals, gamma-ray detection, low noise
Gamma-ray detection by semiconductors and scintillation with large volume andhigh energy resolution has been widely used in medical applications, nuclear
plant monitoring, and archaeology. Due to the strong penetrating ability of
gamma-ray photons, semiconductors for gamma-ray detection should consistof high Z elements, a millimeter-scale thickness, a large carrier mobility-lifetime
product, a low defect density and other strict criteria.
Recently, halide lead perovskite single crystals have been thought of as nextgeneration semiconductors for gamma-ray detection due to their high atomic
number element, large carrier mobility-lifetime product and strong defect tolerance values. However, most solution-processed halide lead perovskite single
crystals suffer from large noise when applied with high bias.
In this abstract, the reasons of the noise is found that for halide lead perovskite
single crystals with large thickness (>1mm)， the noise is more likely come fromthe defects of the bulk crystals instead of the injecting charges. Thus, Schottky
diode based devices could suppress the noise from applied bias, but cannot
suppress the noise from defects.
Consequently, we design some kinds of structures gamma-ray photodiode madeof halide perovskite single crystals including N-I-N type and cascade PIN photodiodes. The N-I-N type photodiode could sufficiently suppress the hole defects existed in the bulk crystals and collect the electrons with high collection
efficiency. Thus, an high energy resolution of 7% @667keV gamma-ray photons
is achieved.
The cascade PIN photodiodes could have a much larger mobility-lifetime productwith structure of MAPbCl3/MAPbBr2.5Cl0.5/MAPbBr3. And the cascade PIN pho-109

Abstracts - Oral

O3 - 02

todiodes could suppress the noise from electrodes. As a result, the cascade PIN
photodiodes shows energy resolution of 4.5%@667keV gamma-ray photons.
Acknowledgement
Acknowledgement: This work is financially supported by National Key Researchand Development Program of China(2021YFE0105900), the South Africa/China
Joint Research Program (UID: 132796)

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#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

Low-noise and High sensitivity direct-conversion X-ray detectors
based on perovskite single-crystal heterojunctions

Yuzhu Pan1, Xin Wang1, Yubin Xu1, Omolola Esther Fayemi2, Qing Li1, Ying
Zhu3, Wei Lei1,*
Joint International Research Laboratory of Information Display and Visualization,
School of Electronic Science and Engineering, Southeast University,
Nanjing/China1, Department of Chemistry, School of Mathematics and PhysicalSciences Faculty of Natural and Agricultural Sciences North-West University,
Mafikeng Campus, Private Bag X2046, Mmabatho 2735/South Africa2, E-Xray
Electronic Co. Ltd., Suzhou/China3
Email: lw@seu.edu.cn
Keywords: X-ray detectors, PIN diodes, epitaxial growth, low noise
X-ray detectors have been actively sought for broad applications in the fields of
security, defense, medical imaging, industrial material inspection, nuclear power
stations and scientific research. Especially, direct X-ray detectors using
semiconductor materials to absorb X-ray photons and create electron–hole pairs
are more compelling owing to its high photoelectric conversion efficiency andhigh intrinsic spatial resolution. Various materials have been exploited for direct
X-ray detection including stabilized amorphous Se (a-Se), PbI2, HgI2, CdTe and
CdZnTe. However, these materials suffer from limitations like high cost,mechanical stiffness, limited detection area, high operating voltage, lowabsorption of X-ray radiation with photon energies >40 keV, and difficulty in upscaling to larger wafers.
Very recently, organic–inorganic metal hybrid perovskites (MHPs), which havebeen shown to be excellent optoelectronic materials for direct X-ray detectors,
combine many of the preferred intrinsic properties for X-ray detection, including
a large proper bandgap (Eg) of 1.6–3.0 eV to reduce thermal noise, a large μτ
product on the order of ∼10−2 cm2 V−1 due to the small electron effective
mass and extremely low trap density, the high-Z element Pb in the compound.
In addition, the simple low-temperature synthesis of its single crystals (PSCs)
give rise to the possibility of fabricating high-performance detectors and forminga large-area flat-panel detector array at low cost.
Due to the strong penetration ability of hard X-ray photons, a large electrical field
was required to drift the photoinduced charges for high-efficiency chargecollection inside thick PSCs. However, the dark current increases with the applied

111

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O3 - 03bias in PSCs-based photodiodes. Meanwhile, the noise current increases andfloods the weak signal current induced by X-ray photons. Herein, we proposed
an approach of fabricating PSCs-based heterojunction by liquid phase epitaxial
growth to form sufficient blocking barriers. Through energy structure design,lattice-matched n-type MHPs (bismuth-doped) and p-type MHPs (silver-doped)
can form as electron/hole blocking layers for sufficient charges barriers.
The fabricated X-ray PDs exhibit a low leakage dark current density of ~ 11 nA
cm−2 and noise current of 47 fA Hz−0.5 (1 mm2, over 500 V cm−1). The large
external electrical field enables a state-of-art response speed (fall) of 750 ns,
large X-ray detection sensitivity of 36.0 μC mGy−1cm−2 and the lowest
detectable dose rate of 16 nGys−1(40 kVp). This work will motivate newstrategies to fabricate high-performance devices based on perovskites using
solution-processed methods. These founding also explore a new generation of
low dose and high dynamic X-ray detectors based on MHPs.
Acknowledgement
This work is financially supported by National Key Research and Development
Program of China(2021YFE0105900), the South Africa/China Joint Research
Program (UID: 132796)

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#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

Application of Unfolding code MEALU to Neutron Spectrum
Measurement using Multi-cylindrical Bonner Counter

Tetsuo Iguchi1,*, Shigetaka Maeda2
Nagoya University/Japan1, Japan Atomic Energy Agency/Japan2
Email: t-iguchi@energy.nagoya-u.ac.jp
Keywords: unfolding code, neutron spectrum measurement, multi-cylindrical
Bonner counterWe developed the new unfolding code, the Maximum Entropy and Maximum Likelihood Unfolding Code (MEALU), based on the maximum likelihood method combined with the maximum entropy method, which can determine a neutron spectrum without requiring an initial guess spectrum. Normal or Poisson distributions
can be used for statistical distributions. MEALU can treat full covariance data fora measured detector response and response function. This code was successfully verified its validity through the neutron spectrum measurement problem using the multiple activation foil method[1].
In this study, we applied MEALU to neutron spectrum measurements using a
multi- cylindrical Bonner counter, which is a typical underdetermined problem inradiation measurements and has more smooth response than multiple activation
foils. The measurements were carried out at several points in and around thereactor top pit room of the experimental fast reactor Joyo MK-II at a reactor
power of 48 MWt, using six Bonner counters and a neutron rem-counter (Studsvik
2202D). The detector response function of the Bonner counter was calculatedusing the Monte Carlo code MCNP and was verified in the standard neutron field.
Neutron spectra were obtained from the measurement results of the six Bonnercounters without using the initial estimated spectra. The obtained neutron spectra
reproduced the measured results well.
To confirm the validity of the unfolding results, they were compared with shielding
calculation results around the Joyo reactor vessel by the transport calculation
code DORT and the well-known unfolding code SAND-II.
The spectral shapes of MEALU and SAND-II in the region below 1 MeV are ingood agreement. Above 1 MeV should be the rest of the high-energy componentof the fission spectrum, and MEALU shows a more plausible spectral shape. Thiscan also be confirmed by comparison with DORT, where the calculation results
tended to overestimate MEALU and SAND-II.
Based on these results, we will discuss the applicability of the MEALU code toneutron spectral unfolding using a multi-cylindrical Bonner counter.
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Reference[1] S. Maeda and T. Iguchi, “A new unfolding code combining maximum entropy
and maximum likelihood for neutron spectrum measurement,” J. Nucl. Sci. andTech. 50(4), p.381 (2013).

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#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

Development of an in-house PMT voltage divider for the TDCR
counting system

Agung Agusbudiman1,*, Kyoung Beom Lee, Jong Man Lee, Sang Hoon Hwang,
Byoung Chul Kim
Department of Precision Measurement, University of Science and Technology(UST)/ Korea1, Radioactivity Metrology Team, Korea Research Institute ofStandards and Science (KRISS)/ Korea2
Email: agung@ust.ac.kr
Keywords: PMT voltage divider, TDCR counting, PMT test
An in-house PMT voltage divider for the Hamamatsu R329-02 PMT used in a
triple-to-double coincidence ratio (TDCR) counting system was developed at
KRISS. The voltage divider was designed based on the Hamamatsu H7195 assembly type with some modifications to meet the requirement of the KRISS-TDCR
system.
The in-house voltage divider was designed using a grounded cathode and tapered scheme with two BNC output ports and three SHV ports to accommodate
the requirement of the TDCR counting experiment. One output port is taken fromthe 10th dynode and used for spectroscopy purposes; the other output is takenfrom the anode and used for counting purposes. The three SHV ports were used
to supply the operating, focus grid, and first dynode high voltage independently.
The variation of voltage applied to the focus grid is necessary to obtain efficiency
variation in the TDCR counting.
This work will present the evaluation of the in-house PMT voltage divider. The
voltage divider was evaluated by operating the Hamamatsu R329-02 PMT in adark box using the picosecond laser diode driven with the PDL800-D from
PicoQuant as the light source. The laser was triggered externally by a waveformgenerator with the lowest frequency of 10 Hz.
For the dynode output, we evaluated the capability of the PMT in detecting single
photoelectron events using a charge-sensitive preamplifier and a GammaVision
spectroscopy system. The PMT linearity was also tested by varying the laser intensity level.
For the anode output, we evaluated the count rates by counting the events above
the threshold, which was determined just below the single photoelectron peak.
The counting was also performed with the variation of the applied voltage of the
focus grid and first dynode to obtain the efficiency variation of the TDCR counting.
The result from the evaluation was then compared with the result obtained fromthe Hamamatsu H7195 assembly type.
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Abstracts - Oral

O3 - 06#3 Radiation Detection and Sensor Technology

Study on FPD for high-energy X-ray CT for fuel debris screening

Akira Mikami1, Yuki Mitsuya1, Hiroyuki Takahashi1,*
University of Tokyo1, University of Tokyo2, University of Tokyo3
Email: leo@n.t.u-tokyo.ac.jp
Keywords: fuel debris, X-ray, GOS,
After the severe accidents of the Fukushima Daiichi Nuclear Power Station(1F).
The fuel debris was created because of the meltdown. The fuel debris mainlycomposed of U, Pu from nuclear fuel, Fe, Zr from structural materials andcladding tube. After extracting the fuel debris from the 1F building, analyzing thematerials of the fuel debris is required for effective storage.
Analyzing the heavy metals such as fuel debris which contains U,Pu, high-energy
X-ray is necessary. For stopping the high-energy X-ray, improvement of
FPD(Flat Panel Detector) is essential. One method is to increase the depth of
scintillator. In our group, a pixelated GOS scintillator with 1.5 mm thickness wasdeveloped with Toshiba corporation. GOS stands for Gd2O2S:Tb. This is realizedwith ceramic GOS which enables to pile up with 1.5mm. For an X-ray source, a3.95-MeV LINAC X-ray source whose maximum energy was 3.95 MeV was used.
With 20-mm Pb pre-hardening filter, the peak energy of X-ray is estimated asaround 1.0 MeV with GEANT4 simulation. For simulated material of fuel debris at
1F, the mixture of Fe, Zr, and Pb was used.
For comparison, XRD 1622 made with Perkin Elmer FPD was used. Its GOS
thickness was 436 μm. The FPD developed in our group had 10 cm × 10 cm, on
the other hand, the Perkin Elmer FPD is 40 cm × 40 cm. Another difference isthat the FPD developed in our group uses the IGZO-TFT, on the other hand,
Perkin Elmer FPD uses the amorphous silicon TFT.
Firstly, we compared the spatial resolution of these two FPDs using resolution
gauges. And then, to measure the features of the FPD, the S/N ratio, the contrastof CT images was compared. The shape of X-ray beam is cone beam. Contrast
and substance identification, FPD developed in our group is better. More analysis
with many conditions to features of FPDs is ongoing.

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O3 - 07#3 Radiation Detection and Sensor Technology

Feasibility Verification of Deep Learning Based 4π Field of View
Collimator-less Imaging System Using Voxelated Single Volume
GAGG(Ce) scintillator: A Monte-Carlo simulation

Ajin Jo1, Dongmyeong Hong2, Wonho Lee3,*
Health Science Research Center, Korea University/Korea1, Department of Healthand Safety Convergence Science, Korea University/Korea2, School of Healthand Environmental Science, Korea University/Korea3
Email: wonhol@korea.ac.kr
Keywords: Deep Learning, Collimator-less Imaging, Voxelated Scintillator,
Monte-Carlo Simulation
Monitoring environmental radiation using source-tracing imaging systems is
expected to be effective for swift response to various situations. Various radiation
imaging systems, therefore, have been developed for environmental radiation
monitoring. A Coded aperture system which uses coded patterned-mask toreconstruct source images is a widely-used system for radiation source imaging.
However, its requirement of the coded-mask limits field of view (FOV) of thesystem. This system is also difficult to be applied for high-energy sources owing
to its characteristics to use photoelectric effects. A Compton camera system
which uses Compton scattering angles of the interactions to trace original source
position can be an alternative system to coded aperture systems contributed toits unlimited FOV and feasibility in high energy region. However, it is difficult toobtain sufficient detection efficiency for development of real-time imaging
system because it uses only Compton scattering interactions detected in different
pixels or voxels of the detector system.
A collimator-less imaging system can be effective system to improve detection
efficiency without FOV limitations. Maximum-likelihood expectation maximization(MLEM) method can be used for source image reconstruction of the collimatorless imaging system using photoelectric absorption or Compton scatteringinteractions detected in the detector. However, the MLEM method requires anumber of iterative operations for each energy bins and image pixels and alsouses only a single interaction type during image reconstruction process, whichleads to longer image acquisition time.
In this study, we suggested a deep-learning (DL) based collimator-less imaging
system using voxelated single-volume GAGG(Ce) scintillator to improve117

Abstracts - Oral

O3 - 07detection efficiency and to reduce calculation time. Intensities of depositedenergy of all detected interactions regardless of interaction types in eachdetector voxel show different patterns depending on the source position. These
patterns were expected to make deep learning models possible to learn andpredict original source position. The feasibility of this DL-based collimator-lessimaging system was verified using Monte Carlo N-particle Transport version 6(MCNP6) code.
A single volume GAGG(Ce) scintillator detector divided by 4×4×4 voxels wasmodeled by the MCNP code for training and test data acquisition. The size of thescintillator and each voxel was 1.2×1.2×1.2 cm3 and 0.3×0.3×0.3 cm3. Theenergy resolution of the detector was set to be 4.9 % at 662 keV and 6.5 % at
356 keV using Gaussian energy broadening (GEB) card. Ba-133 sources wereplaced on the points 20 cm far from the detector center along with constantintervals along the azimuthal and zenithal angles. Data from each source position
were acquired with F6 tally to acquire the amount of deposited energy from all
detected interactions.
A fully-convolutional network (FCN) generally used for image segmentation
studies was composed with Tensorflow 2.0 library for source position prediction.
The training dataset were composed of 1,191 single source data points acquired
from MCNP simulation and 3,000 dual source data points generated by selecting
and averaging two single-source data. Each datapoint were labeled along withthe source position. The full with half maximum (FWHM) of each label was 10
degree to verify the performance of the system. Test dataset was composed of
200 dual-source data which was not used for the training. The performance ofthe system was evaluated with image similarity that shows numbers close to 1 asimages are similar to each other. As a result of DL model training, the FWHM of
the images predicted from single-source data showed 13 degree which is similar
to the values labeled. The image similarity value of the model estimated withdual-source test dataset showed 0.9742.

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#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

Demonstration of Gamma Source Finding in the Field based on
Fusion Data between 4π Gamma Imaging and Simultaneous
Localization and Mapping

Minato Kanda1, Kenji Shimazoe2, Yusuke Tamura3, Hanwool Woo4, Mizuki
Uenomachi5, Fumihiko Ishida6, Jun Kawarabayashi7, Ken'ichi Tsuchiya8, Kei
Kamada9, Hideki Tomita1,*
Graduate school of Engineering, Nagoya University/Japan1, Graduate school ofEngineering, The University of Tokyo/Japan2, Graduate school of Engineering,Tohoku University/Japan3, Faculty of Engineering, Kogakuin University/Japan4,
Center for the Promotion of Interdisciplinary Education and Research, Kyoto
University/Japan5, Department of Electrical and Control Systems Engineering,
National Institute of Technology, Toyama College/Japan6, Faculty of Scienceand Engineering , Tokyo City University/Japan7, Second Department of Forensic
Science, National Research Institute of Police Science/Japan8, New Industry
Creation Hatchery Center, Tohoku University/Japan9
Email: h-tomita@energy.nagoya-u.ac.jp
Keywords: Sensor fusion, SLAM, Gamma imaging
The radiation source finding method is required to prevent terrorist acts using
radioactive source(s) and to find lost or stolen radiation source(s). For operators’safety, it is desirable to use an unmanned system with remote operation. Therefore, we are developing a prototype system using a 4π gamma imager and a 3DLiDAR mounted on an unmanned vehicle[1,2]. Point gamma source(s) can befound based on fusion data from 4π gamma imaging and 3D-LiDAR Simultaneous Localization and Mapping (SLAM). Using the prototype system including a
CdTe pixel array detector as the 4π gamma imager, a field experiment was conducted to find a high-intensity 137Cs gamma point source (2.1×102 MBq) located in a 80 m square field. The prototype system was moved from the start
position (22.7 m far from the source) step-by-step to get closer to the source
direction following the 4π gamma image measured at each position. Finally, thesource position was determined with an uncertainty of several tens of centimeters
and the source intensity was fairly under-estimated due to ignoring the shieldingof the source.
Acknowledgement
This work was partly supported by JSPS KAKENHI Grant Number 19H00881.

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Reference[1] Atsushi Mukai, Hideki Tomita, Shintaro Hara, Keita Yamagishi, Ryohei Terabayashi, Tetsuya Shimoyama, Kenji Shimazoe, Proceedings of 2021 IEEE/SICE
International Symposium on System Integration (SII), 105 – 109, (2021) DOI：
10.1109/IEEECONF49454.2021.9382624 [2] Hideki Tomita, Shintaro Hara, Atsushi Mukai, Keita Yamagishi, Hidetake Ebi, Kenji Shimazoe, Yusuke Tamura,Hanwool Woo, Hiroyuki Takahashi, Hajime Asama, Fumihiko Ishida, Eiji Takada,Jun Kawarabayashi, Kosuke Tanabe, Kei Kamada, Sensors 22, 4325 (2022) DOI：
10.3390/s22124325

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#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

Research on Detection Efficiency Calibration by GEANT4
Simulation for the HPGe Detector in HTR-PM

Rui Nie1, Yu Wang1, Ziling Zhou1, Qian Ma2, Xi Chen2, Zihao Fan3, Juncheng
Liang3, Feng Xie1,*, Jianzhu Cao1
Institute of Nuclear and New Energy Technology, Collaborative Innovation
Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced
Reactor Engineering and Safety of Ministry of Education, Tsinghua University,
Beijing, China1, Huaneng Shandong Shidao Bay Nuclear Power Company LTD,
Rongcheng，Shandong , China2, National Institute of Metrology, Beijing, China3
Email: fxie@tsinghua.edu.cn
Keywords: High-Temperature gas-cooled Reactor Pebble-bed Module;
Radioactive Source Term; High-purity Germanium Gamma Spectrometer;
Detection Efficiency Calibration; Monte Carlo; Geant4
The world's first High-Temperature gas-cooled Reactor Pebble-bed Module(HTR-PM) with the safety features of Generation IV advanced nuclear energy
system was first connected to the grid on December 21, 2021, at Shidao Bay,
Rongcheng, Shandong Province, China. The radioactivity in the primary circuit
can reflect the performance of the fuel element cladding particles in the core andis important for the radiation safety evaluation of HTR-PM. The radioactivity
monitoring for the primary circuit of HTR-PM adopts the combination of online
total γ continuous measurement by NaI detector and nuclide analysis of thehelium sampling from primary circuit by the high-purity germanium (HPGe) γ
spectrometer. The precise determination of types and activities of theradionuclides using HPGe detectors is the basis for reliable and accurate analysisof the radioactive source items in the primary circuit of HTR-PM. In order toquantitatively determine the activities of radionuclides in the sample, thedetection efficiency of the HPGe detection system needs to be calibrated. In thisstudy, Monte Carlo simulation method was used to calibrate the detection
efficiency of the HPGe detection system in the HTR-PM radiochemical laboratory
by GEometry ANd Tracking (Geant4) code. And, the detection efficiency bysimulation was compared with the detection efficiency by experiment using thestandard source of the source term in the helium of the primary circuit. Then, thedifferences between the simulated detection efficiency and the experimental
detection efficiency were also analyzed. It was found that the error of thepreliminary simulation calculation increases with the decrease of γ-ray energy.

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After adjusting the detector model parameters, the error of the detection
efficiency by simulation and the detection efficiency by experiment could becontrolled within 6%. After optimization of the model, the final detection
efficiency curve of the HPGe detector γ-spectrometer of the experimental system
was determined by Geant4. This study can provide a research basis for theaccurate determination of the radioactivity source items in the primary circuit of
HTR-PM, and also provide valuable empirical references for determining thedetection efficiency of the experimental system by using Monte Carlo simulation.

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#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

High-energy photon counting based on Se-doped MAPbI3 single
crystals

Yubing Xu1, Xin Wang1, Shunjie Chai1, Yuzhu Pan1, Omolola Fayemi2, Qing Li1,
Ying Zhu3, Wei Lei1,*
School of Electronic Science and Engineering, Joint International ResearchLaboratory of Information Display and Visualization, Southeast University,
Nanjing, China1, Department of Chemistry, School of Mathematics and PhysicalSciences Faculty of Natural and Agricultural Sciences North-West University,
Mafikeng Campus, Private Bag X2046, Mmabatho 2735, South Africa2, E-Xray
Electronic Co. Ltd., Suzhou 215131, P. R. China3
Email: lw@seu.edu.cn
Keywords: Se-doped,MAPbI3 single crystals,photon counting
The unclear radiation detection, especially gamma-ray spectroscopy, has shownhuge prospect on astrophysics, nuclear material detection and medical treatment.
And solid-state semiconductors based unclear radiation detectors have attracted
lots of attentions. The detection principle is based on high-energy photon
interaction with heavy atoms in the semiconducting materials. In recent years,
the organic-inorganic lead halide perovskites have attracted many attentions forunclear radiation detection due to their high attenuation coefficient, lager
mobility-lifetime product, wide bandgap and low-cost solution growth method.
Additionally, unclear radiation detectors based on MAPbBr3, MAPbBr3-xClx and
MAPbI3 single crystals with schottky junctions have been reported for highenergy photons counting in last five years. Compared to detectors with schottky
junctions, detectors with pn junctions show more effectively blocking of injectedcurrent with barriers and more quickly separating the electrons and holes pairs
with built–in electric field. Mostly, p-n photodiodes are fabricated with organic
materials, carbon derivatives and nano-oxides as blocking layers by spincoating method, vapor-phase epitaxy and ink jet printing. However, theunavoidable lattice mismatches between PSCs substrates and blocking layers
lead to the poor quality of blocking layers and defects exist in the interfaces,
which cause the leakage of dark currents and noise currents.
In our previous work, Se-doped MAPbI3 single crystal have pn junctions formed
during crystallization process.
Herein, we fabricated a device based on Se-doped MAPbI3 single crystals grownthrough inverse temperature growth method for 241Am α-particle photon

123Abstracts - Oral

O3 - 10counting.
Benefit from this self-formed pn junction, our devices have low defects density,
which decreases the leakage currents and noise currents under high working
voltages. By directly converting the single photon signal into an electrical pulse
signal, our device based on Se-doped MAPbI3 single crystals can acquired
spectra241Am α-particle. In conclusion, Our devices have fairly low dark current
density of 4.14 10-7 A/cm2 under 400 V/cm. Furthermore, this device with pnjunctions based on Se-doped MAPbI3 single crystals acquired 241Am α-particle
spectra with a resolution of 47.2%.
Acknowledgement
This work is financially supported by National Key Research and Development
Program of China(2021YFE0105900), the South Africa/China Joint Research
Program (UID: 132796)

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O3 - 11#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

Ultra-dense Ceramic Scintillators for Positron Emission
Tomography

Sun Il Kwon1,*, Yimin Wang2, Minseok Yi3, Urmila Shirwadkar2, Steven Onorato2,
Junghune Nam2, Jae Sung Lee3, Jarek Glodo2, Kanai Shah2, Simon Cherry1
Department of Biomedical Engineering, University of California Davis, Davis, CA
95616, USA1, Radiation Monitoring Devices, Inc., 44 Hunt St., Watertown, MA
02472, USA2, Interdisciplinary Program in Bioengineering, College ofEngineering, Seoul National University, Seoul, 03080, South Korea3
Email: sunkwon@ucdavis.edu
Keywords: Ceramic scintillator, lutetium oxide
Positron emission tomography (PET) is one of the major biomedical imaging
modalities and is widely used to diagnose diseases, especially cancer. A PET
scanner is built with a number of gamma detectors to capture 511-keV
annihilation photons, and each gamma detector consists of scintillators,
photosensors, and electronics. To detect more gamma photons, scintillators withhigh density and stopping power scintillators are preferred. Most commercial PET
scanners employ lutetium oxyorthosilicate (LSO) or bismuth germanate (BGO)
scintillators that have favored properties for PET, such as high density (7.4 g/cm3
for LSO and 7.1 g/cm3 for BGO). This presentation will introduce noveltransparent ceramic scintillators based on lutetium oxide (Lu2O3). In comparison
to current scintillators used in PET scanners, these ceramic scintillators have an
extremely high density (9.4 g/cm3) and stopping power. Thus, scintillator arrays
used in a PET scanner can be made thinner without sacrificing detection
efficiency compared to the current PET scintillators. Simulation studies showedthat 15 mm of these new scintillators can achieve detection efficiency equivalent
to 20 mm of LSO. The reduction in thickness will also improve energy and timing
performance and reduce cost and depth-of-interaction effect. Importantly, these
new scintillators have a cubic crystal structure that allows fabrication using
ceramic processing in place of conventional single crystal growth, with thepotential for a significant reduction in cost compared with LSO. We havefabricated and evaluated several Lu2O3-based ceramic scintillators with different
combinations of (co-)dopants. Yb3+ doped Lu2O3 has an exceptionally fast
decay time (~1.7 ns). For the first time, we have observed that codoping
substantially increases the light yield by introducing a slower component. By
tuning codopant concentrations, we can optimize brightness or pulse shape as125

Abstracts - Oral

O3 - 11desired, depending on the ultimate application. The properties and initial
evaluation of the new scintillators appear promising for PET applications.
Moreover, these new ultra-dense ceramic scintillators have strong potential toextend to other areas of radiation detection.
Acknowledgement
This research was supported by NIH grants R01 EB030538 and R44 CA243762.

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O3 - 12#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

Development of Glass Gas Electron Multiplier based Beta Ray
Imaging System

Moh HAMDAN1, Yuki MITSUYA2, Kenji SHIMAZOE1, Hiroyuki TAKAHASHI1,*
Department of Nuclear Engineering and Management, Graduate School ofEngineering, The University of Tokyo1, Institute of Engineering Innovation,
School of Engineering, The University of Tokyo2
Email: leo@n.t.u-tokyo.ac.jp
Keywords: Micro Pattern Gas Detector, Glass Gas Electron Multiplier, Beta Ray
Imaging
Radiation imaging technology is widely used in various fields of science including
agriculture. Beta particles have now been widely utilized to determine
physiological phenomena in living plants through autoradiography techniques.
On the other hand, the need for a real-time imaging process, the ease ofobservation in bright environments, the stable measurement for long periods oftime, the use of low-energy radiation, and large area capability are challenges inthe selection or even development of imaging detectors for living plant
autoradiography.
Micro Pattern Gas Detectors (MPGDs) as imaging detectors have been widely
utilized especially for X-ray imaging given their typically high position resolution
capabilities. One of the MPGDs that is widely used is the Gas Electron Multiplier(GEM). In this work, a thick gas electron multiplier with a glass substrate called
Glass Gas Electron Multiplier (G-GEM) was developed as a beta ray imager
considering its capability to meet the challenges of detector development for the
autoradiography process.
G-GEMs were fabricated using Photosensitive Etchable Glass (PEG3 or PEG3C)
substrates using photolithography and etching techniques. They have sensitive
areas of 100 × 100 mm2. The thickness, hole size, and hole pitch are 680 µm,
170 µm, and 280 µm, respectively. G-GEM was placed inside the gas chamber
between the aluminized Kapton cathode and charge readout anode with gaps of
5 mm and 3 mm, respectively. The chamber was supplied with Ar/CH4 (90/10)
gas with a flow rate of 20 mL/min and a pressure of 0.1 MPa. The effective gainand energy resolution of G-GEM were first characterized by using a plate anodeusing 5.9 keV X-ray radiation from a 55Fe radiation source. Furthermore, a beta
ray imaging demonstration was conducted utilizing a collimated 90Sr beta ray
source placed at various positions of the chamber window at an effective gain127

Abstracts - Oral

O3 - 12of around 6000. A charge-division imaging scheme utilizing a continuous
resistive plate of Indium Tin Oxide (ITO), as well as individual readout, isdeveloped and tested in turn in this section. The counting system consisting of
charge sensitive preamplifier, shaping amplifier, dynamic Time Over Threshold(dToT) signal processing, and FPGA was developed.
The effective gain and energy resolution were characterized based on the mean
peak of 55Fe energy spectra. The effective gain of G-GEM achieved 47000
without cascade assembly. This value is highly desirable, especially for lowenergy imaging applications. The optimum energy resolutions were around 18%
and at Full Width at Half Maximum (FWHM). The two-dimensional distribution ofbeta-ray radiation can be observed as a result of demonstration imaging. These
results show that the development of G-GEM detectors is promising for an
application utilizing beta particle emitting sources such as living plant
autoradiography.

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#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

Dosimetric Study of a Novel Perovskite Scintillator for HighEnergy Radiotherapy

Ning Gao1, Zongming Huang2, Jieping Zhou4, Yihan Zhang3, Xi Pei5, Zhengguo
Xiao2, Xie George Xu1,*
School of Nuclear Science and Technology, University of Science andTechnology of China, Hefei, China1, Hefei National Laboratory for PhysicalScience at the Microscale, Department of Physics, CAS Key Laboratory ofStrongly-Coupled Quantum Matter Physics, University of Science andTechnology of China, Hefei, China2, Institute of Energy, Hefei Comprehensive
National Science Center, Hefei, China3, Department of Radiation Oncology, The
First Affiliated Hospital of University of Science and Technology of China,
Hefei, China4, Technology Development Department, Anhui Wisdom Technology
Company Limited, Hefei, China5
Email: xgxu@ustc.edu.cn
Keywords: Radiation detector, Perovskite, Scintillator, Radiotherapy,
Radiotherapy dosimetry
PurposePerovskite scintillators are the latest class of radiation detectors, yet theirpotential for radiotherapy dosimetry is not fully investigated. This study aims tocharacterize dosimetric properties of thin films based on perovskite as
scintillators to detect high-energy radiation in radiotherapy.
Methods
The scintillator film was formed by mixing perovskite scintillator powder withpolymer and was coated on a glass substrate. The film was radiated by a medical
linear accelerator to characterize its dosimetric properties in high-energy photon
and electron beams. The emitted light was captured using a CCD digital camera.
For the dose-response, the scintillator film was irradiated over a range of 20–500
MU. Intensity of images were plotted versus MU. For dose rate linearity, the dose
rate is varied by changing the photon beam over the range 100–600 MU/min. For
relative light-yield measurements, a CsI:Tl scintillator was used as the reference.
In addition, energy dependence, sensitivity, and radiation tolerance were also
tested, respectively.
Results129

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The radioluminescence (RL) intensity of the film demonstrated an outstanding
linear response to radiation dose for 6 MV photon and 6 MeV electron beams.
The film showed linearity between the X-ray dose rate and RL intensity. Afterbeing radiated with high radiation dose, the film remained great sensitivity. More
important, the RL intensity of the film remains relatively constant (within 3%) overa wide range of doses. Moreover, the film showed no significant energy
dependence for photon beams of 6 MV, 6 MV flattening filter-free (FFF), 10 MV,
and 15 MV. The light-yield measured was higher than conventional CsI:Tlscintillator.
Conclusion
In this study, perovskite-based thin films were developed. We found that
proposed films have properties that make it promising for clinical dosimetry inradiotherapy. As a result, they have the potential to be used for quality assurance
in radiotherapy. Furthermore, they show promise as flexible detectors.

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#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

Background Reduction in PET using Quantum Entanglement
Property of Annihilation Gamma Rays

Donghwan Kim1,*, Kenji Shimazoe1, Hiroyuki Takahashi1
Department of Nuclear Engineering and Management, Graduate School ofEngineering, The University of Tokyo, Japan1

Email: kim-donghwan@g.ecc.u-tokyo.ac.jp
Keywords: PET, Background Reduction, Quantum Entanglement
Positron Emission Tomography (PET) is a radiation emission imaging modality
used for functional imaging in nuclear medicine. PET is mostly used for cancer
imaging with 18F-FDG but other cancers were imaged with other isotopes suchas 124I imaging for thyroid cancer, 68Ga PSMA prostate cancer, and 86Y for
dosimetry information prior to 90Y, etc. In addition to random and scatter coincidence backgrounds observed in pure β^+ emitters, these non-pure isotopes
suffer from background caused by prompt gamma rays. Also, PET is used forhadron therapy range verification in which case in-beam monitoring suffers frombackground event.
Gamma rays emitted from positron-electron annihilation are entangled and polarized orthogonally to each other. It results in a strong correlation in the difference between azimuthal Compton scattering angles of the annihilation gamma
rays. The Double Compton Scattering (DCS) cross section of entangled and polarized annihilation gamma rays describes the modulation by azimuthal scattering
angle difference.
From the equation, it can be known that if data is acquired with both polar scatterangles around 81.2°, the contrast between 0° and 90° of Δϕ becomes the largest.
Therefore, taking data with 90° of Δϕ and polar scatter angles around 81.2°, datawith higher true fraction can be obtained compared to normal PET data acquisition.
Experiments were conducted with an in-house PET system composed of GAGG
scintillation crystals. 8×8 crystals were assembled for one array of crystal which
is attached to the same size of SiPM arrays. In order to measure DCS and subsequent absorption events within detector module, all scintillator and SiPM pairs
were processed parallelly and independently.
System modelling included all the four crystal locations and their energy depo-131

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sitions to accurately model the response. Each crystal was divided into subcrystals and all possible combinations of the rays connecting sub-crystals wereconsidered for projection in the imaging space and energy depositions in the
crystals. In addition, background estimation on each Line of Response (LOR) andedge preservation prior were implemented in image reconstruction for lower noiseand lower background imaging.
18F-FDG was used for measurements. Signal to Background Ratio (SBR) evaluation of small vial source measurement showed that it can reduce the background by approximately two times while image became noisier. Spill over ratio
evaluation from cold rod phantom showed that it improved from 7 % to 5 % andbackground was also reduced by approximately two times.

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#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

Radionuclide analysis Using PHITS and Neutron Irradiation in
HANARO Concrete Sample

Shin Dong Hyuk1, Kim Jong woo1, Kown Jin Hyung 1, Kim Han Jin 1, Kim Young
Soo 2, Kim Yong Kyun 1, Kim Ji Seok 2,*
Department of Nuclear Engineering, Hanyang University, Seoul, Korea1,
HANARO Utilization Division,Korea Atomic Energy Research Instittute, Daejeon,
Korea2
Email: geesuck@kaeri.re.kr
Keywords: Neutron Activation Analysis, Concrete, PHITS
Radiated concrete accounts for about 40% of the cost of decommissioning
nuclear power plants. The parent nuclide of Co-60 and Eu-152 are evaluated as
trace elements of several ppm in the material, but account for more than 90% of
the total dose when decommissioning nuclear facilities. Therefore, in order toreduce the cost of decommissioning nuclear facilities, it is necessary to collectinformation on parent nuclides in materials and establish an efficient dismantlingplan based on accurate assessment of radionuclide inventory.
Radionuclides in the sample were determined by comparing the calculated value
using the PHITS code with the measured value in the neutron irradiation
experiment using the research reactor. The concrete material to be measured
was collected by drilling a cylindrical hole in the outer wall of the containmentbuilding of the HANARO research reactor of the Korea Atomic Energy ResearchInstitute. In order to ensure uniformity of the collected sample, it was pulverized
into a powder form using a ball mill. Neutron irradiation and measurement
conditions are as follows. For the experiment, distribute about 0.1g of eachconcrete sample to a polyethylene container. This was investigated for 1, 20, 40,
and 60 minutes with an output of 15 MW in the PTS #1 irradiation process, andthe thermal neutron at this time is about 1.28E+13n/cm2·sec. The gamma-ray
measurement of the radiated concrete sample was performed using the HPGe
detector for about 5 months and about 80000 seconds of count time. The
calculation conditions using the PHITS Code set the neutron irradiation timeequally, and it is expected that there will be differences in the material
composition and the neutron energy spectrum.
Radionuclides such as Sc-46, Co-60, Zn-65, and Eu-152 could be detected in
the gamma spectrum obtained through the energy calibration process. Later, it
was confirmed that the total count was high in the order of Zn, Sc, Co, and Eu,

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although efficiency calibration using certified standard materials was required. In
the PHITS code simulated using the material components of the NUREC report,
Mn-54 was detected in addition to the aforementioned nuclear species, which is
expected to be caused by differences in the neutron energy spectrum. In addition,
the intensity of radioactivity also differed from the experimental value, which isbelieved to be due to the difference in concentration of impurities.
Through this study, it was possible to analyze the components of trace elements
in concrete.At the time of decommissioning, the doses of Co-60 and Eu-152 aredominant, but Sc-46 and Zn-65 account for a high proportion of the total doses
in about one year after permanent shutdown. If the material components of eachnuclear facility are analyzed through experiments and calculations, it can be used
as an input value to determine radionuclide inventory.
Reference
1. IAEA, "Radiological Characterization of Shut Down Nuclear Reactors for
Decommissioning Purposes," Technical Report Series No. 389(1998) 2.
NUREG/CR-3474, “Long Lived Activation Products in Reactor Materials,”(1984)134

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#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

Development of Prototype Backscatter X-ray Imaging System for
Security Screening

Geunyoung AN1, Junsung PARK1, Hee SEO1,*, Byeong-no LEE2
Jeonbuk National University, Jeonju, Korea1, Korea Atomic Energy Research Institute, Jeongeup, Korea2
Email: hseo@jbnu.ac.kr
Keywords: Security screening, Backscatter X-ray imaging, Design optimization,
and Prototype system
In recent years, there has been a rapid increase in trade volume, leading to anincrease in attempts to smuggle illicit goods into the country. In response, theKorea Customs Service (KCS) has implemented an X-ray imaging technique toinspect all incoming items, thereby detecting illegal and dangerous goods andprotecting the safety of citizens. Nevertheless, the present system for inspecting
small cargo relies on transmission X-ray imaging, making it difficult to discernorganic and powder-form materials in the images.
Backscatter X-ray imaging is highly sensitive to materials with low atomic numbers because of the large probability of Compton scattering. Consequently, it has
the potential to increase security screening personnel’s detection efficiency for
identifying illicit organic compounds like explosives, drugs, and seeds. In thisstudy, a pencil-beam-based backscatter X-ray detection system was considered, which offers superior beam intensity per image pixel over other backscattermethods.
In the present study, we optimized the design of a backscatter X-ray detection
system for luggage screening and estimated its performance using the Monte
Carlo simulation technique. The system consists of an X-ray generator, a rotatable collimator (i.e., chopper wheel), and large-area detectors, each of which
was optimized in terms of tube voltage, beam resolution, and detection efficiency,
respectively. The performance of the optimized system was estimated under various object conditions (i.e., detector-to-object distance, object thicknesses,
and material compositions) as well as for postal phantoms.
Based on the optimized design, we developed a prototype backscatter X-ray
scanning system and evaluated its performance at various tube voltages for different objects. It was found that the isolation contrast, which indicates the minimum discernible thickness of an object when the background material differsfrom the object to be imaged, was approximately 1 mm. Moreover, contraband135

Abstracts - Oral

O3 - 16

items randomly concealed within the luggage were imaged, resulting in clearlyvisible images at tube voltages ranging from 80 to 160 kVp. It is expected thatthe backscatter X-ray security scanner can provide enhanced detection efficiency for thin objects and/or organic materials.

136

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#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

Development of Background Simulation Technique for Radiation
Portal Monitors at Airports and Ports in South Korea

Hayoung SIM1, Hyosang CHOI1, Hee SEO1,*, Hyeonjun CHOI2, Haneul KIM2,
Jaeho JANG2
Department of Applied Plasma and Quantum Beam Engineering, Jeonbuk
National University/Korea1, Department of Radiation Monitoring at Ports ofEntry, Korea Institute of Nuclear Safety/Korea2
Email: hseo@jbnu.ac.kr
Keywords: Radiation Portal Monitor (RPM), Monte Carlo simulation, Sitespecific Background Model, Suppression
At domestic ports and airports, radiation portal monitors (RPMs) have been
installed and operated for detection of illegal radioactive materials. Commercial
RPM systems are based on a large polyvinyl toluene (PVT)-based scintillation
detector. In order to detect radioactive materials passing through the RPM, thealarm criterion is set based on the background count rate, which is the sitespecific factor depending on the operating location due to compositiondifferences in soil and rocks, weather variation (i.e., rainfall and temperature) as
well as self-shielding effect by container vehicles. Therefore, it is essential toevaluate the effects of background radiation variation according to various
factors for consistent detection capability with the optimized RPM’s alarm level.
In this study, we established a procedure of developing a background simulation
model of commercial RPMs using Monte Carlo simulations, which could be thebest option for considering various conditions as an alternative to field tests on
24/7 operating sites. The developed background simulation model, based on
which we were able to apply the realistic characteristics of RPM sites at airports
and ports in South Korea, was validate by comparing measured and simulated
background spectrum, and then a site-specific source term library forbackground radiation was constructed using Surface Source Write and Read
(SSW/SSR) techniques to improve computational efficiency. Additionally, theeffects of rainfall and temperature on the background count of commercial RPMs
(models: 4525-3800 and 7000, Ludlum), which are operated at the Incheon andDonghae ports in Korea, were analyzed using a 3-year database of RPM
background signals and a database of rainfall and temperature provided by the
Korea Meteorological Administration (KMA). In the rainfall-effect analysis, it was
found that the averaged background count during rainy events increased by up

137

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to approximately 20% as a function of the amount of rainfall, depending on theregion wherein each RPM was installed. In the case of temperature, thebackground counts on sunny days varied within ~4.7% within the temperature
range of -5℃ to 30℃.
Finally, the variations in the background count caused by various conditions of
the cargo container (i.e., cargo type, loading amount, and payload area) were
evaluated. To this end, the cargo container vehicle with a size of 40 ft, the mostcommonly used in Korea, was precisely modeled. The simulated backgroundcount for the empty container vehicle were compared with the measured data,
recorded at 0.2-second intervals during the vehicle passes between the RPM, tovalidate the simulation models. When the empty container passed through theRPM, the average background count decreased by approximately 20–30% with aconsistent tendency. For various cargoes (i.e., potassium chloride, slag powder,and zircon), as the vertical loading increased, the background radiation varied byapproximately 50–60% on average, and up to 74% compared to before the vehicle
entered the RPM. The degree and shape of background suppression varied
depending on the cargo conditions.
The results of this study can be used to optimize the site-specific alarm criterion
for a RPM by realistically estimating the background count variation due to various
factors.

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O3 - 18#3 Radiation Detection and Sensor Technology

Performance Estimation of PVT Scintillator-based Seafood
Monitoring System for Screening of Radioactive Materials

Hyosang CHOI1, Hee SEO1,*
Jeonbuk National University, Jeonju, Korea 1
Email: hseo@jbnu.ac.kr
Keywords: Seafood, Radioactive material, Screening, Monte Carlo
In order to secure safe food for the public, the Republic of Korea has establishedstandards for the management of radioactive materials in food. Notice No. 202313 of the Ministry of Food and Drug Safety presents the regulatory criteria forradioactive materials in seafood as 100 Bq/kg for Cs-134+Cs-137 and I-131,
respectively. The current method based on gamma-ray spectroscopy using ahigh-resolution HPGe detector has a limitation in that rapid inspection cannot be
performed because of a time-consuming pretreatment process as well as a longmeasurement time (typically 10,000 s). Therefore, we designed a polyvinyl
toluene (PVT) plastic scintillator-based monitoring system using Monte Carlosimulations that can determine compliance with the regulatory criteria on a nearreal-time basis. Additionally, the minimum detectable activity (MDA) of three
types of seafood (i.e., dried laver, raw oyster, and fishcake) for Cs-137 and I131 were calculated to estimate the performance.
To determine the optimal thickness of the PVT detector, the detection efficiency
was estimated by varying the detector thickness from 1 to 10 cm with 1-cm
intervals using the MCNP code. The background count rate should be minimized
to reduce the MDA; therefore, the lead shield was applied to all surfaces of thedetector except for the detection window that faced the seafood. The thicknessof the lead shield was changed from 0 to 2.5 cm with 0.5-cm intervals and thebackground count rate with the lead shield was calculated by the MCNP code. In
order to determine the MDA of Cs-137 and I-131 for each seafood, thebackground count and detection efficiency were calculated. To this end, theregion of interest (ROI) around the Compton edge was set to reduce the influenceof the natural background count and electronic noise. Finally, the MDA of Cs137 and I-131 was calculated for the dried laver, raw oyster, and fishcake, whichis representative of seafood for seaweed, shellfish, and fish.

139

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As a result, the optimal thickness of the PVT detector was determined to be 5 cm
providing reasonable detection efficiency. Furthermore, the thickness of the leadshield was optimized to be 0.5 cm as the background count rate decreased by
up to 50% using 0.5 cm of lead shielding compared to without shielding. Finally,
the MDA for Cs-137 was determined to be 53.8, 68.4, and 58.2 Bq/kg in thedried laver, raw oyster, and fish cake, respectively. Additionally, the MDA for I131 was determined to be 47.1, 52.4, and 43.7 Bq/kg for each seafood.
In this study, it was found that the designed monitoring system could detect theradioactive materials in seafood with an MDA of less than 100 Bq/kg, which is
the regulatory criteria. For the next step, the prototype system with theradioactivity analysis software will be developed.

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Radiation Hardness Assurance Testing of a 40-nm
Microprocessor Applied in Nuclear Industry

Runcheng Liang1, Faguo Chen1,*, Rong Guo1, Zhirui Zheng1, Ri Zhao1, Jing
Zhang1, Zhaoxing Liu1, Xing Liu1
China Institute for Radiation Protection1
Email: chenfaguo1@126.com
Keywords: total ionizing dose effects, radiation damage, microprocessor
The integrated chips' gamma irradiation damage (total ionizing dose effects, TID
effects) is the main bottleneck for applying electronic equipment in the gamma
radiation environment. As the technology nodes of integrated circuits reach thenanometer level, reduction of the gate oxide thickness and changes of the fieldoxide isolation technology lead to a change in the main influencing factors of the
TID effects and requires further in-depth research. This paper explores the TID
effects failure mode and the discrete nature of the failure dose for a 40 nm
commercial microprocessor that has been widely applied in the nuclear industry.
We developed an extendable microprocessor TID effects online test system bycombining electrical parameter measurement and functional parameter check.
The system has 16 test methods covering six main functions of the device,including the kernel, embedded memory, digital-analog/analog-digital signal
processing, clock, device power consumption, and communication functions.
For a 40 nm microprocessor manufactured by STMicroelectronics, the TID failuremode under dynamic bias conditions has been measured by the proposed test
system. After the irradiation dose reached 377.44 ± 20.34 Gy(Si), all samples
experienced digital communication interruption, a sudden drop in current
consumption, and abnormal ADC/DAC output. In contrast, no memory checkerrors occurred during the irradiation. Based on all 16 kinds of parametricmeasurement or functional check results, the TID effects of the tested
microprocessor were probable to be a functional failure caused by some kernel
instructions. Besides, the neutron-gamma synergistic damage has been
investigated. The samples under test were first irradiated with 1E8/cm2 neutrons
and then placed into a Co-60 radiation field for gamma radiation. It has beenfound that with the mixed neutron-gamma irradiation, the failure dose of the
microprocessors was 409.17±26.75 Gy(Si), which has increased by 8.4%
compared with the result of the gamma irradiation experiment. It is argued thatneutron irradiation produces vacancies in bulk silicon that form double vacancies

141

Abstracts - Oral

O3 - 19and stable defect centers when combined with impurity atoms. Gamma irradiation
produces vacancies that would integrate with these defect centers or point
defects/defect clusters produced by neutron irradiation, reducing the defectdensity and resulting in milder TID effects.
For the discrete nature of the failure dose for the 40 nm microprocessor, thestatistical analysis showed that the failure does obey the logistic distribution. Dueto the part-to-part variability of TID effects, the SMOTE-Bootstrap algorithm has
been developed for fitting the failure distribution model with an extremely small
sample size (nIn summary, an extendable microprocessor TID effects online test system has
been established, and the TID effects failer dose of a commercial 40 nm
microprocessor has been measured. A statistical analysis method with based on
the SMOTE-Bootstrap algorithm has been proposed. It has been proven to besuited for the evaluation of the failure dose distribution with a sample size smaller
than 10.

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O3 - 20#3 Radiation Detection and Sensor Technology

Scintillation Properties of Cr-doped α-Al2O3 Crystals for theDecommissioning Monitoring System

Shunsuke Kurosawa1,*, Akihiro Yamaji2, Daisuke Matsukura2, Akira Yoshikawa2
Tohoku University & Osaka University1, Tohoku University2
Email: kurosawa@imr.tohoku.ac.jp
Keywords: Scintillator, Radiation Dose, High Dose Rate
To remove the debris in the Fukushima Daiichi Nuclear Power Plant in thedecommissioning step, real-time dose monitors are required to be used under
the high dose-rate condition We have developed such dose-rate monitor
consisting of a scintillator, optical fiber and CCD spectrometer, and scintillation
photons are read with the CCD under the lower dose condition through the fiber.
Here, the length of the fiber is 20 -200 m. Under the high dose rate (over a few
Sv/h), the optical fiber itself emits scintillation photons as noise originating fromsome defect or OH bonding in the fiber, and such noises are dominant below
550 nm. Moreover, Cherenkov photons are also generated in the fiber with a main
emission band below 400 nm. Scintillation materials with longer emission
wavelengths of over 600 nm have merit to avoid such fiber noises, and also highlight-output materials are required to have good sensitivity.
Cr-doped α-Al2O3, ruby, is known as red and infrared emission scintillationmaterial with a sharp peak at 694 nm, while this material has long persistence.
To suppress such long persistent emissions, we optimized the concentration of
Cr in this paper. The crystals with the different Cr concentrations from 0.1 to 2%
were grown by the micro pulling down method, and these samples wereconfirmed to have a single crystal phase using a powder X-ray diffraction pattern.
Additionally, some optical properties such as transmittance spectra and emission
measurements.
long persistent for these samples were evaluated excited by X-rays and gammarays. The long persistence for the 1%Cr-doped α-Al2O3 was relatively
suppressed up to now compared to other Cr-concentration. We show the details
and discuss the mechanism of the long persistence in this paper.

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Research on the performance of SiC thermal neutron detector
based on metal-oxide-semiconductor structure

Lei Ren1, Yuncheng Han1,*, Jiacheng Zhang1, Rui Zhang1, Houjun He1, Xiaoyu
Wang1
Hefei Institute of Physical Science, Chinese Academy of Sciences/China1
Email: yuncheng_han@163.com
Keywords: SiC; thermal neutron detector; metal-oxide-semiconductor structure;energy resolution; neutron detection efficiency
Semiconductor-based neutron detection technology provide a critical technology for radiation protection. The Schottky barrier diode (SBD) is widely utilized insemiconductor detector due to its simple structure and thin dead layer, but its
energy resolution is constrained by high leakage current. Metal-oxide-semiconductor (MOS) based semiconductor detectors offer a promising solution to improve energy resolution by mitigating the reverse leakage current and enhancing
the homogeneity of the Schottky barrier. The MOS neutron detector is composedof neutron conversion material (NCM), gate electrode, oxide layer, semiconductorand back electrode. Increasing the thickness of NCM improves the probability ofneutron reaction, which enhances the neutron detection efficiency (ηd). Increasing the thickness of the oxide layer weakens the tunneling effect in the MOS
device, thereby reducing the reverse leakage current. However, thicker NCM andoxide layer lead to increased energy deposition by secondary charged particles,
degrading energy resolution. Therefore, optimizing the thickness of both NCM
and oxide is essential to enhance detector's performance. SiC as a typical widebandgap semiconductor has shown potential for applications in extreme environments, such as high temperature and hard radiation.
This paper aims to conduct research on the optimization design and key preparation processes of SiC-based MOS devices. The main approach for optimization
design is based on the Monte Carlo program. Using MCNP to calculate the impactof NCM types and geometric parameters on ηd. Additionally, the energy spectrum
broadening caused by the energy deposition of secondary charged particles in
the NCM, gate electrode, and oxide layer is calculated to determine the energy
resolution of the device. The electrical properties of SiC-based MOS devices aretested after fabrication. The SiO2 thickness is obtained through high-frequency
C-V testing. The linear energy transfer of secondary charged particles in the SiC
epitaxial layer is calculated using SRIM, and the charge collection efficiency of144

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the device under different biases is obtained. The leakage current of the device
is obtained through I-V testing, and the optimal operating voltage and corresponding characteristic peak energy resolution can be determined. 10B NCM
films are prepared according to the optimization results. The neutron responseof MOS devices is intended to be performed in 241Am-Be or D-T neutron source
to obtain the ηd and energy resolution of characteristic peak.
Preliminary analysis of testing results indicates that the MOS device exhibits areduction in leakage current by over two orders of magnitude compared to the
SBD device at a bias voltage of 100 V, providing technical support for the development of high-performance new neutron detectors.
Acknowledgement
This study was supported by Anhui Provincial Key R&D Program (202104g010
2007), and Collaborative Innovation Program of Hefei Science Center, Chinese
Academy of Sciences(2022HSC-CIP024).
Reference[1] Karadavut O, Chaudhuri S, Kleppinger, et al. Performance-Improved Vertical
Ni/SiO₂/4H-SiC Metal–Oxide–Semiconductor Capacitors for High-Resolution
Radiation Detection. IEEE Transactions on Nuclear Science, 2022, 69(8): 19651971. [2] Shi D, Wang Y, Wu X, et al. Improving the barrier inhomogeneity of
4H-SiC Schottky diodes by inserting Al2O3 interface layer. Solid-State Electronics, 2021, 180: 107992. [3] Jia Y, Shen Y, Sun X, et al. Improved performanceof SiC radiation detector based on metal–insulator-semiconductor structures.
Nuclear Instruments and Methods in Physics Research Section A Accelerators
Spectrometers Detectors and Associated Equipment, 2021, 997(3): 165166.

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Proposal of Method for Estimation of Gamma Source Direction
based on Self-shielding Profile in a Multipixel Array Detector

Atsushi Mukai1, Minato Kanda2, Hideki Tomita2,*
Department of Energy Engineering, Graduate School of Engineering, Nagoya
University, Japan1, Department of Aplied Energy, Graduate School ofEngineering, Nagoya University, Japan2
Email: h-tomita@energy.nagoya-u.ac.jp
Keywords: Radiation Ditection, Gamma-ray Source Identification, Neural
Network
We propose a method to estimate gamma-ray incident direction to a multipixel
array detector. When gamma rays are emitted from a point source incident into
the detector, more events occur at pixels closer to the source and fewer eventsoccur at pixels further away from the source. The direction of the source can be
estimated by analysing the spatial distribution of counts in the detector, nameda “self-shielding profile”. A prediction model based on the self-shielding profile
was developed to estimate the direction of the gamma source using a
convolutional neural network. To verify the accuracy of the model, source
direction estimation experiments were conducted. The direction of a 137Cs point
source (1.8 MBq) was varied around a CdTe pixel array detector with 30×12×4
pixels. As a result of estimating the source direction for the acquired data, 59%of all data were estimated to have an angular difference of fewer than 10 degrees
from the true source direction.
Acknowledgement
This work was partly supported by JSPS KAKENHI Grant Number 19H00881.

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#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

Fuel debris mapping tool using an Unmanned Underwater Vehicle
to grasp fuel debris submerged in Fukushima Daiichi Nuclear
Power Plant

So Kamada1,*, Tetsuichi Kishishita2, Takahiro Oyama2, Kenichi Terashima3, Keisuke Okumura2, Kazuya Nishimura3
National Maritime Research Institute1, High Energy Accelerator Research Organization2, Japan Atomic Energy Agency3
Email: kamada-s@m.mpat.go.jp
Keywords: fuel debris, neutron, UUV
To accelerate the decommissioning of the Fukushima Daiichi Nuclear Power Plant(1F), we developed a fuel debris mapping tool to grasp the distribution of submerged fuel debris. The tool comprises neutron sensors, a camera, sonar, and
LiDAR and is incorporated into an unmanned underwater vehicle.to grasp thesituation inside the submerged fuel debris in unit 1 and unit 3 of 1F. The positioning system of UUV’s position in the reactor utilizing the Inertial Navigation
System (INS) was also developed because the Global Positioning System (GPS)cannot be used in the submerged environment. The experiment was conductedin a robot test tank at the Naraha Center for Remote Control Technology Development of the Japan Atomic Energy Agency.
The neutron sensor can discriminate and measure neutron signals under a highdose gamma-ray environment (~10 Gy/h) by combining a semiconductor neutron sensor and a dedicated radiation-resistant signal processing circuit. In addition, a bubble detector, which is insensitive to gamma rays, is installed as apassive neutron sensor, and neutron information can be obtained efficiently by
combining this information.
On the other hand, by combining sonar, cameras, and LiDAR, it is possible tovisually grasp the situation inside the reactor. In LiDAR measurement, since there
is no ghost image mixture due to laser, walls, floors, surrounding structural materials, and multiple scattering, it is possible to acquire 3D point cloud data of
the situation inside the reactor with high accuracy. Combining the count information of these neutron detectors and the visual conditions obtained by the underwater search devices inside the reactor significantly contributed to the efficient exploration of submerged fuel debris.
To obtain the UUV’s position in the reactor, the Inertial Measurement Unit (IMU)
comprises a gyro sensor and an acceleration sensor was used in the INS positioning system without the GPS. The signal obtained with the IMU was integrated
twice to derive the position of the UUV.
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O3 - 24#3 Radiation Detection and Sensor Technology

Fabrication of Transition Edge Sensors with faster response time

Takeshi Jodoi*, Yuki Mitsuya, Ryan Smith, Amin Choghadi, Hiroyuki Takahashi
Department of Engineering, The University of Tokyo1
Email: jodoi-takeshi602@g.ecc.u-tokyo.ac.jp
Keywords: superconductor, microfabrication,photon-number-resolving
With the development of information society, computer plays an important role
to process huge amount of information. For that reason, the higher performancecomputer is desired to be realized, but it is difficult for classical computer tomake a dramatic progress. Therefore, the quantum computer which has a different operating principle attracts people’s attention.
To realize optical quantum computer, which is worked by optics, it is required todevelop a photon-number-resolving detector with small time jitter in the nearinfrared or communication wavelength. Transition Edge Sensors (TESs) arepromising photon number resolving detectors for near infrared photons, which
can realize high energy resolution, low energy threshold, a high quantum efficiency, and a low dark count rate. However, the response of TESs is very slow,
and we need to improve its response speed to sub-nanosecond level. To realize
fast response TESs, we should make the sensitive area of TESs smaller. The timerequired for the thermalization will be faster with smaller devices.
We simulated the response time of 1 μm \times 1 μm TES at 1500 nm wavelength.
This simulation shows that 1 μm TES has faster than nano second and can realize
fast response characteristics. In this research, we will fabricate TESs with small
sensitive area and investigate the characteristics of TESs.
We fabricated this small TESs by using photolithography and lift-off. In this process, the edge of sensitive area has burr and this interrupt the contact between
sensitive area and Nb electrode. This causes the large resistance and interrupt
the TESs to become superconducting state under transition temperature. So wewill use FIB (Focused Ion Beam) to delete the burr in the sensitive area andimprove the connection between sensitive area and electrode.
In addition to this, we formed Si wafer into a racket-like shape which matches
the inner diameter of optical fiber in order to improve the alignment between
optical fiber and the sensitive area of TESs.
We will show you the result of fabrication and the characteristics of the devices.
Acknowledgement
This work was partially supporte by JST Moonshot R&D Grand No.JPMJMS2064-2

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#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

Muography-based method of subsidence risk assessment forunderground tunnel construction

Naoya okamoto1,*, Shoichiro Kawase1, Ayumu Okuda1, Hiroto Shiba1, Yukinobu
Watanabe1
IGSGS, Kyushu University/Japan 1
Email: okamoto.naoya.515@s.kyushu-u.ac.jp
Keywords: Muography, subsidence, underground cavity
The shield tunneling method is a common technique used in the construction oflarge underground tunnels. Although it is considered relatively safe, subsidence
accidents caused by this method have recently occurred. To prevent such
accidents, it is necessary to detect and fill underground cavities that are formed
during excavation before a collapse occurs. However, existing methods for
exploring underground cavities, such as ground-penetrating radar, have difficulty
detecting cavities located deeper than 10 m from the ground.
In this situation, we propose a method of subsidence risk assessment bymonitoring underground cavities generated during tunnel excavation using
muography, i.e., a non-invasive exploration method that utilizes cosmic-ray
muons. By measuring the muon flux penetrating the object being explored, we
can investigate its internal density structure like a computer tomography (CT)scan. One advantage of muography over CT scans is the high penetrating powerof cosmic-ray muons, which allows monitoring of the upper bedrock even fromunderground tunnels that are several tens of meters long.
The core of our proposed method is a muon tracker moving along with tunnel
excavation to measure the underground muon flux. It in real-time while drilling,
it is possible to detect cavities created by excavation. This is because the flux ofmuons coming from the direction of the cavity increases compared to the caseof no cavity.
This paper is devoted to two main tasks. The first one is a simulation-based
feasibility study exploring underground cavities with muography. We used a
Monte Carlo simulation and an empirical formula to calculate the muon flux in thesubsurface. PHITS[1] was used in the Monte Carlo simulation and Miyake's
empirical formula[2] was used in the empirical equation. From the result,149

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we confirmed that a spherical underground cavity with a diameter of 2.5 m whose
center resides 5 m below the surface from 30 m underground, can be statisticallysignificantly discriminated in a one-hour measurement. The second one is thedesign of the muon tracker. The design of the muon tracker was determined bysimulation results and required performance. The detailed simulation results andthe design of the muon tracker will be presented at the symposium.
Reference[1] S. Miyake, J. Phys. Soc. Japan, 18, 1093 (1963) [2] T. Sato, et al., J. Nucl.
Sci. Technol. 55(5-6), 684-690 (2018)150

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#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

Waveform discrimination by fitting derivative of synchronized ideal
normalized curves, dSINC fit

Martin Andersson1,*, Tatsuo Torii2, Susumu Ryufuku1, Ryohei Kurosawa1, Hiroko
Kido1
Visible Information Center, Inc., Tokai-mura, Japan1, Institute of Environmental
Radioactivity, Fukushima University, Japan2
Email: martin@vic.co.jp
Keywords: waveform-discrimination, peaks-piling, triple-layer phoswich, KNN,
computer code, numerical solution, thorium
A detector that combines two or more scintillators with different optical properties
and an optical-to-electrical signal conversion device is called a phoswich (orphosphor sandwich) detector. Phoswich detectors are commonly used forradiation detection where the type of incidence radiation needs to be determined.
In phoswich detectors, scintillator materials with different optical properties, e.g.
gain and decay times are selected and combined so that the shape of the signal
pulse output from the conversion device depends on the relative contribution of
scintillation light from two or more scintillators. Feature extraction andsubsequent KNN, is the most commonly used method of discriminating the
waveform (WF)output to derive the particle type. In the KNN method, the feature
extraction process can be sensitive to noise and overlapping waveforms (socalled peaks-piling) which affects the accuracy of the both the training andclassification phase.
As an alternative to the KNN method, we are developing the DSINC (derivative ofSynchronized Ideal Normalized Curves (dSINC) method. dSINC solves some ofthese feature-extraction problems by sidestepping them and fitting the derivativeof the entire gain (or head) portion of the waveform against a set of normalized
ideal waveforms learned from the training-data.
In our previous work [1] we showed the successful training process to determine
the 'ideal waveforms', and an improvement in discrimination accuracy ofwaveforms using dSINC with these 'ideal waveforms' compared to KNN on adataset from our triple layer phoswich detector. Further dSINC allowed for theinspection of the fitted data against the ideal waveforms for verification.

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We plan to improve the usefulness of this dSINC method by analyzing actual
measurement data and establish it as a general-purpose αβγ discrimination
method. We working on improving the intuitiveness of the Confidence Scoringmechanism. And whereas the original dSINC tests were performed on individual
waveform data where each individual peak was gathered and treated separately,
we have now adapted the algorithm to work efficiently on sequential datastreams. We are also working on further reducing the sensitivity to noise, by
optimizing the noise reduction algorithm used in the original work. While dSINC
in its original form is less sensitive than KNN to distortions caused by peakspiling, the combination of having access to ‘Ideal Waveforms', and working on
the gain (leading edge timewise) of the waveform allows us some exciting new
possibilities of solving the ideal subcomponents of combined (piled) waveforms
by the cancelation of ideal waveforms from the real data in the input-buffer. We
also attempt to illustrate how peaks-piling affects KNN vs dSINC.
Reference
M. Andersson, et al. Waveform discrimination by fitting derivative of synchronizedideal normalized curves, dSINC fit (preprint),
https://doi.org/10.48550/arXiv.2212.10716

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Development of Dual Energy X-ray Security Scanner for Improved
Organic Material Detection: A Monte Carlo Study

Junsung Park, Geunyoung An, Hee Seo*
Department of Applied Plasma and Quantum Beam Engineering, Jeonbuk National University, Republic of Korea1
Email: hseo@jbnu.ac.kr
Keywords: X-ray security scanner, Variable tube voltage, Organic materials detection
The X-ray security scanner, which uses a transmission X-ray imaging technique,is applied to detect contraband, including drugs and agricultural products. Current X-ray security scanners use fixed tube voltage, typically 160 kVp. The highpenetrating power of X-rays is favorable to detect high-density and/or highatomic number materials (e.g., guns and knives); however, it limits the detectionof organic materials. Especially, there is a need to improve organic material detection (i.e., drugs and explosives), which are low-density powdered organic
materials. Lower tube voltage affords an advantage in terms of improved contrast
for organic materials at the cost of penetrating power. In this context, it would
be beneficial to develop an X-ray security scanner with variable tube voltages
optimized for the object to be inspected (e.g., higher tube voltage for higher
density materials and vice versa).
In the present study, by Monte Carlo simulations, we designed an X-ray security
scanner with variable tube voltages. The dual-energy detector and X-ray generator were optimized in terms of various parameters. First, the optimized design
for radiation generator with tungsten target was determined in terms of target
thickness and angle. The source term was obtained for the designed radiation
generator (i.e., 80–160 kVp). Then, the source terms were used to determine theoptimized design of the dual-energy detector in terms of scintillator materials,
thickness, and a low-energy filter. Finally, we developed a simulation model for
X-ray security scanners to evaluate their performance.
The reliability of the X-ray security scanner simulation model was validated bycomparing the image profiles of the simulated and experimental images. The
experimental setup for the source-to-detector distance and source-to-object
distance was 70 cm and 120 cm, respectively. Simulations were performed under
the same conditions, and the results were in good agreement (i.e., within lessthan 5%) for pixel counts. Therefore, we concluded that our simulation model

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would be useful in obtaining realistic images of an X-ray security scanner.
The performance of X-ray security scanner with variable tube voltages was evaluated using a simple phantom composed of organic materials with similar properties (i.e., density and atomic ratio). The phantom was constructed using cylindrical high-density polyethylene (HDPE) and a contrasting material, such as water. Images were obtained using a 3 mm thick phantom at 80 and 160 kVp source
terms. In the 80 kVp image, HDPE could be visually distinguished from water incontrast to the 160 kVp image. The contrast was determined to be more thandoubled, 1.7 ⅹ 10-3 and 4.6 ⅹ 10-3 for the 160 and 80 kVp images, respectively.
In the present study, we developed an X-ray security scanner simulation modelwith variable tube voltages and evaluated its performance in detecting organic
materials using a simple phantom. Compared to conventional X-ray security
scanners, we have demonstrated a significant improvement in the detection oforganic materials. Our study concluded that an X-ray security scanner with variable tube voltage could be used to effectively detect organic materials, such as
drugs and explosives.

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#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

Characteristic evaluation of gamma-ray spectrometer using an
unfolding method with refined response functions

Masateru Hayashi1,*, Makoto Sasano1, Tetsushi Azuma2, Taisuke Makita1, Masakazu Nakanishi2, Hiroshi Nishizawa3
Advanced Technology R&D Center, Mitsubishi Electric Corporation/Japan1, Energy Systems Center, Mitsubishi Electric Corporation/Japan2, Department ofApplied Nuclear Technology, Fukui University of Technology/Japan3
Email: Hayashi.Masateru@ct.MitsubishiElectric.co.jp
Keywords: gamma-ray spectroscopy, scintillator, response functions, unfolding,uncertainty
In case of a radiation-leakage accident at a nuclear power plant or radiationrelated facilities, it is necessary to measure the energy spectrum of gamma rays
to identify the radioisotopes for evaluating the external and internal exposure
dose. Although high-purity germanium (HPGe) semiconductor detectors areused for gamma-ray spectrometry due to their high energy resolution, they havea high initial and operational cost due to the requirement of a cooling mechanism.
On the other hand, Thallium-doped sodium iodide (NaI(Tl)) scintillators, althoughhaving difficulties with radioisotope identification due to their poor energy resolution, have the advantage of improving sensitivity and decreasing measurement
time compared with HPGe semiconductor detectors due to the easiness of crystal
enlargement and without any cooling requirement. Although several unfolding
methods have been studied to improve the energy resolution of scintillators, conventional methods have mainly been used to evaluate environmental dose rates
and are not suitable for radioisotope identification because the discriminable energy difference is more than 50 keV. Thus, it is possible to achieve a low-costand easy-to-handle gamma-ray spectrometer using NaI(Tl) scintillator by improving the unfolding method. Therefore, we improved the unfolding method witha refined response function that considered the energy dependence of scintillation efficiency for secondary electrons produced by the interaction with the incident gamma rays. Response functions were calculated using a Monte-Carlo simulation code EGS5 with a mesh width of 5 keV in the energy range of 0 to 3 MeV,
assuming gamma rays irradiated parallel to the sides of the detector. The unfolding algorithm was an iterative approximation method, which is independentof the initial guess. As a result of repeated measurements using Cs-137, Co-60,
Ba-133, Na-22, and Y-88 radiation sources, the energy resolution is about 10

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keV and the measurement accuracy of gamma-ray fluence rate is almost constant at less than 10% at 0.3 MeV and above. Also, we confirmed the correlation
between the uncertainty of the unfolding result and the statistical uncertainty ofthe measured energy spectrum. As a result, the uncertainty of the unfolding resultin a single measurement has been evaluated.

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#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

Multi-Class Radioisotope Identification using Convolutional Neural
Networks with Two-Dimensional Input Data

Yong Hyun Kim1, Yong Kyun Kim1,*, Yang Han Cheol, Kihong Pak, Jae Chang
Kim, Jae Young Jeong
Department of Nuclear Engineering, Hanyang University, Seoul, Korea 047631
Email: ykkim4@hanyang.ac.kr
Keywords: Radioisotope Identification, RIID, Convolutional Neural Networks,
Artificial Neural Network, CsI(Tl), Gamma-ray Spectrum
Artificial Neural Networks (ANNs) have been widely studied and successfully
applied to radioisotope identification (RIID). In addition, recently, Fully Connected
Neural Networks (FCNNs), and convolutional neural networks (CNNs) have been
extensively studied among ANNs. CNNs use fewer parameters owing to weightsharing and sparse connectivity realized through convolutions, which reduces
training time and mitigates overfitting. CNNs assume that the inputs have a local
spatial structure and are trained to extract and identify the best feature such asa photopeak or Compton continua. In a few studies comparing the performanceof CNNs with that of FCNNs and other ML methods for the RIID, CNNs have been
noted to exhibit comparable or higher accuracy in some cases.
In this study, we propose a novel approach to enhance the performance of RIID
using a CsI(Tl) scintillation spectrometer. We reshape 1-D gamma-ray spectraof radioisotopic mixtures into 2-D image data as training data to improve the RIID
performance of the CNN model. Two different datasets are generated, one
consisting of a 1-D composite spectra of 8 radioactive sources, and the otherproduced by reshaping the same 1-D composite spectra into 2-D image data.
Each model is trained and tested with both datasets to predict the relative countdistributions for 8 RI classes, solved as a multi-output regression problem using
softmax and cross-entropy. We have shown through experiments that the use of
categorical cross-entropy effectively minimizes the distance between thepredicted and true contributions. The CNN model trained with the transformed 2D data showed better RIID performance than that trained with 1-D spectral data,with mean magnitude of relative error (MMRE) averages for all RIs of 4.56% and6.28%, respectively. Our results demonstrate that the proposed input
transformation extends the receptive field and enables the model to findmeaningful relationships between spatial features. In conclusion, the CNNs
model trained with the 2-D data is advantageous for the automatic identification

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O3 - 29of multiple peaks that are overlapped in real-time, making it suitable for in-situ
and on-site RIID applications.
Acknowledgement
This work was partly supported by Institute of Information & communications
Technology Planning & Evaluation (IITP) grant funded by the Korea
government(MSIT) (No.2020-0-01373,Artificial Intelligence Graduate School
Program(Hanyang University)) and the research fund of Hanyang University. This
work was supported by the National Research Foundation of Korea(NRF) and the
Korean Association for Radiation Application(KARA) grant funded by the Korea
government(MSIT) (No. 2019M2D3A2060217).

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O3 - 30#3 Radiation Detection and Sensor Technology

Pile-up Separation and Pulse Shape Discrimination for an
Organic Scintillation Detector using Multi-task Learning

Byoungil Jeon1, Junhyeok Kim2, Myungkook Moon1,*
Korea Atomic Energy Research Institute1, Korea Advanced Institute of Scienceand Technology2
Email: moonmk@kaeri.re.kr
Keywords: Pulse shape discrimination, pile-up separation, deep learning,
multi-task learning
Organic scintillation detectors have been considered as alternative materials tohelium, lithium, and boron in neutron detection. In neutron measurement
environments, gamma rays are mostly accompanied by nuclear reactions ofneutrons and environmental materials, and organic scintillation detectors
measure not only neutrons but gamma rays. Therefore, additional techniques todiscriminate n/g signals are necessary to use them in neutron detection.
Fortunately, the shapes of neutron and gamma ray pulses in several organic
scintillation detectors differ from one another due to differences in linear energy
transfer (LET). So classically, their signals can be distinguished using thetechnique called pulse shape discrimination (PSD).
As its name implies, the PSD technique analyzes the shape of pulses. Therefore,
they can’t be applied when pulse shapes are not fully apparent in measured
signals, e.g., piled-up signals measured in high-count-rate environments. In thispresentation, we introduce a deep learning model that conducts pile-up pulse
separation and discrimination of pulse types simultaneously. With the scheme ofmulti-task (MT) learning, we designed a deep learning model that outputs theheight of pulses and the index of radiation types from the pulse signal input. We
generated a dataset using experimental data. We employed a convolutional
neural network as the structure of our model, mean squared error for the pulse
height estimation task, and binary cross-entropy for the n/g discrimination task.
Deep learning models were implemented in the Python environment using the
PyTorch library and trained for 500 epochs using the Adam optimizer. We also
implemented and trained individual models for each task to confirm theeffectiveness of MT learning. We confirmed that the MT model has better
performance on estimation of pulse heights and recall of the neutron pulses in
the generated dataset compared to the individual task models. Furthermore, we
also confirmed that the MT model has higher stability of neutron counting with159

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less signal loss but less error in predicted gamma ray spectra than the classical
PSD method for experimental data.
Acknowledgement
This work was supported by the Ministry of Oceans and Fisheries (KIMST)[Project No. 20200611]

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#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

Application of a novel scintillation gel-based detector for 3D
dosimetry in radiotherapy

Hua Li1, Liang He3, Haijing Jin1, Xuewen Yan1, Hui Zhang2, Deyuan Li2, Liye
Liu1,*
China Institute for Radiation Protection, China1, Shanxi Provincial Key
Laboratory for Translational Nuclear Medicine and Precision Protection, China2,
Department of Engineering Physics, Tsinghua University, China3
Email: liuliye@cirp.org.cn
Keywords: 3D dosimetry, radiotherapy, dose verification, scintillation gel, 3D
optical measurement system
As a critical part of quality assurance and quality control of radiotherapy, pretreatment dose verification would contribute to accurate design and effective
implementation of treatment plan, and could lead to the reduction of dosedelivery error, radiological side effects and tumor recurrence. However, thecurrent dose verification in radiotherapy often adopts low-sampling (such as
point or planar) dose measurement methods, which can not provide high spatial
resolution 3D dose verification in relevant volumes. Therefore, it is difficult tomeet the needs of 3D dose verification for precise dynamic radiotherapy
techniques such as IMRT, VMAT, and SBRT.
In this paper, a novel scintillation gel-based detector was proposed and showed
good potential for 3D dose verification of radiotherapy. The developed novel
scintillation gel material has a photonic soft tissue equivalence (better than 5%),and its luminescence intensity could reach about 80% of a commercial liquid
scintillator (Gold AB) in experimental test. At the same time, the material has
good dose linearity and energy response, high irradiation endurance (>1000 Gy),and no dose rate dependency, as well as other excellent properties suitable fordose measurement. Moreover, the good plasticity of the material makes it apromising choice for functional physical organ phantom.
Then a rapid measurement device for 3D dose distribution in radiotherapy wasdeveloped, including the novel scintillation gel and a 3D optical measurement
system. The hardware of the 3D optical measurement system included cameras
from three orthogonal views, lenses, optical mounts, reflectors, calibration
boards, a light shield, and a computer, among which the cameras and lenses
can be replaced according to the imaging accuracy and imaging distance. The
software of the 3D optical measurement system consisted modules of optical161

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simulation, image acquisition, and data processing. The ZEMAX software wasapplied in the optical simulation module to calculate the system response matrixof the measurement device, describe the propagation process of light from thescintillator to the camera, and form the dose equation. The image acquisition
module was realized by MVS software. The MATLAB-based in-house program
Gel3Ddose was implemented as the data processing module for data loading,image preprocessing, equation solving, detector calibration, and result analysis.
Especially, this paper proposed the LASSO-TV model for dose equation solving,which achieved total variation (TV) minimization, non-negative constraint, andnoise suppression during the iterative process. And detector calibration used thereference radiation field.
The newly developed 3D dose distribution rapid measurement device was thenvalidated in clinical experiments using the Elekta Synergy linear accelerator at the
Shanxi Bethune Hospital. The novel scintillation gel material was irradiated with aclinical treatment plan of stereotactic radiotherapy (SBRT) for intracranial tumor.
And high-resolution measurement of the 3D dose distribution inside thescintillator was achieved. The measurement results were compared and analyzed
with the calculation data of the treatment planning system (TPS) and themeasurement data of the 2D ionization chamber (MapCHECK2). The results
showed that, compared with the TPS, the gamma pass rates (3mm/3%) of theselected coronal plane, transverse plane, and sagittal plane data were all above
90%. Compared with the MapCHECK2 readout, the relative dose error was lessthan 3.5%, which met the error range requirement of 5% for clinical dose
measurement by ICRU. Meanwhile, the novel scintillation gel-based detector also
had high spatial resolution (mm level) and real-time measurement, which wasbeyond the capability of 2D ionization chambers. Such outstanding advantages
made it more suitable for measuring high dose gradient in precise radiotherapy,
and demonstrated great potential in 3D high-resolution dose verification
application for precise dynamic radiotherapy techniques.
Acknowledgement
This research work was supported by the Young Scientists Fund of the National
Natural Science Foundation of China (Grant No.11905190).
Reference
1. Rilling M, Allain G, Thibault S, Archambault L. Tomographic-based 3D
scintillation dosimetry using a three-view plenoptic imaging system. Med Phys.
2020 Aug;47(8):3636–46. 2. Hui C, Robertson D, Beddar S. 3D reconstruction of
scintillation light emission from proton pencil beams using limited viewing
angles—a simulation study. Phys Med Biol. 2014 Aug 21;59(16):4477–92.

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Building and performance test of X-ray communication system
based on grid-controlled modulated X-ray tube and LYSO-SiPM
X-ray receiverJunxu Mu1, Yunpeng Liu1, Xiaobin Tang1,*
Department of Nuclear Science and Technology, Nanjing University of
Aeronautics and Astronautics, China1
Email: tangxiaobin@nuaa.edu.cn
Keywords: X-ray communication, grid-controlled modulated X-ray tube, LYSOSiPM detector, BER
X-ray communication (XCOM) is an optical wireless communication technology
using X-ray as the information carrier. Based on its advantages of high energy
and strong penetration, XCOM has been proved to be able to show uniqueadvantages in solving deep space communication, spacecraft reentry blackout,and Martian dust environments. The existing XCOM can only use intensity
modulation/direct detection (IM/DD) technology to transmit information using thedegree of freedom of X-ray intensity.
We have developed an X-ray signal transmitting and receiving device signal
modulation and demodulation module to realize X-ray communication underlaboratory conditions. We propose a groove-shaped grid-controlled modulated
X-ray tube (GGMXT) with hot cathode It converts the voltage signal loaded on
the grid into the corresponding X-ray signal to realize the conversion from analog
signal to X-ray signal. Based on the non-dominated sorting genetic algorithm II(NSGA-II), with cut-off voltage and tube current as objective function, we
optimize the design of the proposed GGMXT grid structure and provided it withthe characteristics of low cut-off voltage and high emission X-ray intensity. At
the receiving end, an X-ray detection module based on LYSO scintillator coupledsilicon photomultiplier tube (SiPM) is designed. LYSO, as an X-ray detection
material, converts X-rays into visible light, while SiPM converts visible light into
electrical signals. The analog and digital circuits at the back end of the receiverare designed to realize signal shaping, decision and data export. In view of the
problems of LYSO's own radioactivity and the weak X-ray signal in the actual
communication process, we design a time-slot synchronous clock recovery
module. At the same time, we added an active quenching module to obtain an
active quenching (AQ) LYSO SiPM pulse X-ray signal receiver.

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We have built an XCOM system and carried out performance test. The pseudorandom code PRBS7 was used to test the BER, and the signal received by thereceiver was demodulated offline by the lower computer. The experimental results
show that the performance of XCOM system with the reduction of data rate andthe decrease in dead time. The current XCOM system can achieve the data rateof 2.0 Mbit/s at a BER of 3.8e-3.

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#4 Environmental Radiation Measurement and Assessment#4 Environmental Radiation Measurement and Assessment

Characterization of dose- and time-dependent radiation effects
on human skin by multiomics and histological analyses

JiEung Kim1, Won-Suk Song2, Jae-Seung Lee2, Jun Woo Lim2, Ji-Hyeon Park2,
Sang Hyoun Choi3, Dongchan Jang1, Il Won Kim2, Jae Hyun Jeong2, Yun-Gon
Kim2,*
Department of Nuclear and Quantum Engineering, Korea Advanced Institute of
Science and Technology (KAIST)1, Department of Chemical Engineering,
Soongsil University2, Research Team of Radiological Physics andEngineering, Korea Institute of Radiological and Medical Sciences (KIRAMS)3Email: ygkim@ssu.ac.kr
Keywords: ionizing radiation, human skin reaction, multi-omics analysis
Ionizing radiation is widely applied to humans for various medical applications,even though it can cause deterministic or stochastic effects [2]. Apart frommedical exposure, radiation should be always considered because of its
ubiquitous nature. Thus, radiation-induced biological effects are fundamentals
to be studied for the proper use of radiation and our safety.
Skin is the outermost physiological barrier of the human body, and the skin organ
is constituted of the epidermis, dermis, and stratum cornea. When the skin organ
is irradiated, functional damage like desquamation, ulceration, and necrosis
arises, and even oncogenesis can occur [2]. Previous research studied thecellular effects of ionizing radiation on skin cells [3-5], but research with abroader perspective, including cellular scale to organ scale, is essential for betterinterpretation and prediction of real biological effects. In this study, we used bothhuman skin keratinocytes (HEKn), the cells of the skin epidermis, and the skinorganoid, which is the in-vitro 3D full-thickness skin model to investigate thebiological changes in irradiated skin [1].
Human epidermal keratinocyte cells (HEKn) were cultured in cell culture flasks.
3D human reconstituted full-thickness skins (skin organoids) were fabricated.
Then, the cell culture flasks and skin organoids were irradiated with different
doses of 0.1 Gy, 0.5 Gy, and 2 Gy X-ray by a Clinac iX linear accelerator.
Metabolic responses after irradiation on HEKn cells were observed based on themulti-omics analysis. Histological analysis and immunofluorescence imaging
were done for the irradiated skin organoids to observe morphological changes.

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According to the LC-MS/MS results from HEKn cells, irradiation at low doses (0.1
Gy and 0.5 Gy) inhibited the expression levels of functional proteins related toenergy metabolism on Day 0. Conversely, irradiation with 2 Gy increased theproteomic expression of energy metabolism on both Days 0 and 2. In addition,
proteins related to epidermal cornification were significantly upregulated at all
doses on Day 2.
The immunofluorescence staining images of skin organoid at day 14 shows theincrease in the expression level of a constituent in the cornified layer following
irradiation at Day 14. These results suggest that it would be necessary to study
those proteins that were over-expressed related to cornification in this study andtreat them in an initial phase after irradiation to hamper the development of skinpathologies related to epidermal cornification.
In conclusion, this study investigated the dose- and time-dependent changes
after X-ray irradiation based on metabolites from HEKn. Energy metabolism, DNA
repair, and epidermal cornification were induced by the ionizing radiation. In
addition, the observation from skin organoids supported the occurrence of
epidermal cornification. These results provide clues for predicting effects on thehuman epidermis after exposure to ionizing radiation and suggest the possibilityof metabolites as the bio-marker for irradiation.
Acknowledgement
This work was supported by the National Research Foundation of Korea (NRF2017M3A9E4077235, NRF-2018R1D1A1B07048185, NRF-2019M2C8A2058418,
NRF-2020R1A6A1A03044977) and the Korea Health Technology R&D Projectthrough the Korea Health Industry Development Institute, funded by the Ministryof Health & Welfare, Korea (HP20C0210).
Reference
1. Song, et al. "Multiomics characterization of dose- and time-dependent effectsof ionizing radiation on human skin keratinocytes," Korean Journal of Chemical
Engineering, (2022) 2. Stewart, et al. "ICRP publication 118: ICRP statement on
tissue reactions and early and late effects of radiation in normal tissues andorgans–threshold doses for tissue reactions in a radiation protection context."
Annals of the ICRP, (2012) 3. Hu, et al. "Metabolomic response of human skintissue to low dose ionizing radiation." Molecular BioSystems, (2012) 4. Sekihara,
et al. "Low-dose ionizing radiation exposure represses the cell cycle and protein
synthesis pathways in in vitro human primary keratinocytes and U937 cell lines."
Plos one, (2018) 5. Lamartine, et al. "Activation of an energy providing response
in human keratinocytes after γ irradiation." Journal of cellular biochemistry, (2005)

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Development of a measuring method of cosmic-ray muon
momentum distribution using drift chambers

Naoto Nakagami1,*, Satoko Kamei1, Shoichiro Kawase1, Akira Sato2, Satomi
Nakatani1, Yukinobu Watanabe1
Kyushu University, Fukuoka, Japan1, Osaka University, Osaka, Japan2
Email: nakagami.naoto.200@s.kyushu-u.ac.jp
Keywords: cosmic-ray muon, drift chamber, PHITS, soft error
In today’s information society, electronic devices are indispensable and play awide range of roles. Especially in recent years, electronic devices that supportsocial infrastructure, such as automated driving and supercomputers, haveappeared and are expected to become very popular in the future. Therefore,
electronic devices will need to have higher reliability in the future.
One of those that degrade the reliability of electronic devices is soft errors which
are a temporary malfunction of electronic devices caused by radiation incident
on semiconductors. Recently, cosmic-ray muons are attracting attention as oneof environmental radiations that cause soft errors. Cosmic-ray muons arecharged particles that fall from space to the ground and are classified into
positive and negative muons. Previous studies have shown that negative muons
have a larger contribution to the occurrence of soft errors [1]. However, thepreviously measured muon flux data, either with or without charge identification,
are not sufficient to evaluate the soft error rate with high precision.
Thus, we propose a new muon detector system to measure the muon flux datawith charge identification on the ground. It consists of a permanent magnet and
two drift chambers installed above and below the permanent magnet. Cosmicray muons passing through the permanent magnet bend their tracks due to themagnetic field. Since the displacement of the track depends on the charge andmomenta of the muon, they can be identified from the tracks detected in the two
drift chambers.
This paper is devoted to the development of a measuring method of momentum
distribution, which is essential for our proposed measurement system. To
develop the algorithm, muon tracks of various momenta were generated event by
event using PHITS code [2]. Then the muon momentum for each event wasestimated by numerical optimization methods using the classical Runge-Kutta
method and the quasi-Newton method. It was confirmed that the developed
momentum search algorithm was able to estimate muon momentum regardless167

Abstracts - Oral

O4 - 02of the position dependence of magnetic flux density. The details of the developed
algorithm and its accuracy will be presented.
Acknowledgement
This work was supported by JSPS KAKENHI Grant Numbers JP21J12445.
Reference[1] S.Manabe et al., IEEE Trans. Nuclear Science, vol.65, no.8, p.1745-1747(2018). [2] Tatsuhiko Sato et al., Features of Particle and Heavy Ion Transport
code System (PHITS) version 3.02, J. Nucl. Sci. Technol. 55(5-6), 684-690(2018).

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Determination of Ra-226 in Biological Samples
via α-spectrometry Using Ra-223 as Yield Tracer

Ping XU 1, Guobing YU 2, Huiyun XUE 1, Huiping TANG 1, Shiquan CAO1,
Chen ZHI1,*, Yujia KANG1
University of Science and Technology of China, Hefei 230027, China1,
Anhui Radiation Environmental Supervision Station, Hefei 230071, China2
Email: zchen@ustc.edu.cn
Keywords: Ra-226, Ra-223, chromatography, DCyTA, biological samples
The mining and processing of uranium ore will release a large amount of Radium226 (Ra-226) into the environment. Once Ra-226 enters the human body, it willcause continuous internal irradiation to the human body. Therefore, it is of greatsignificance to accurately measure the activity of Ra-226 in biological samples.
The aim of the study was to develop a method to determine Ra-226 in the plant
samples from a uranium mine via α-spectrometry using Ra-223 as yield tracer.
In order to avoid the interference of Ba on the electrodeposition of Ra and theinterference of Th, U, and Ac radioisotopes on the α-spectrum of Ra-226, we
optimized the method to separate Ra using chromatography at first. AG 50W-X8
resin (NH4+ form) was used to purify Ra. Ethylene diamine tetraacetic acid (EDTA)
and 1, 2-diaminocyclohexane-N, N, N', N'-tetraacetic acid (DCyTA) solutions at
various pH were utilized to separate radium from other alkaline-earth cations andTh, U, and Ac, et al. Organics from the AG 50W column were removed in AG 1X8 (Cl- form) resin so that radium can be plated with acceptable yields in anelectrolyte mixture of 0.1 mol·L-1 HCl/ethanol. Then, Ra-226 was determined
via PIPS alpha-detector (ORTEC). The method was applied to the analysis ofadiantum capillus-veneris, which collected around a uranium mine.
Experimental results of the separation of Ra via ion exchange chromatographydemonstrated Ba can be eluted from the AG 50W column in a DCTA 0.075 M
solution at pH 8 without radium loss. Experimental results of determination of
Ra-226 in adiantum capillus-veneris showed that the average activity concentrations of Ra-226 were 924.84 Bq·kg-1 in the leaves and 308.92 Bq·kg-1 in thestems. The chemical recovery yield of samples ranged from 50% to 82%, with anaverage yield of 62.42%. The minimum detectable activity concentration of themethod was 3.66 Bq·kg-1 at the 95% confidence level for 24 h of counting times.
The results obtained through α spectrometry analysis showed good consistency
with γ spectrometry analysis, which need 30 days of measurement time. Alpha
spectrometry analysis has the advantage of analyzing the biological sample,
which does not need to wait for the sample to reach the radioactive equilibrium,and only needs a small amount of sample.
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Numerical analysis and discussion on the implementation of
airborne radioactive emission standards of nuclear power plantradiation protection standard (GB6249-2011)

Fu Shen
CIRP,China1
Email:
Keywords: Airborne radioactivity, emissions, standards, sampling
representativeness, emission stream monitoring
China issued a new nuclear power plant radiation protection standard (GB62492011) in 2011, on the basis of giving the total annual dose limit (0.25mSv/a) for
airborne radioactive emissions (such as chimney emissions), respectively for"particles", "iodine" and other types of emissions in the emission stream, in Table
1 to give a non-nuclear "Bq/a" emission limit regulations, which put forwardcertain requirements for detection capabilities, and there are still many problems
in actual implementation. In this paper, the sampling standard for airborne
radioactive material in the chimney of nuclear power plant (NB/T 20374-2016) iscarried out by numerical modeling and computational analysis. Studies haveshown that monitoring airborne radioactive emissions, such as stack emissions,
not only takes into account nuclides, but also the performance of certain
characteristic parameters such as mixing characteristics for Peugeot can lead tolarge differences in results. This paper puts forward suggestions for similar
problems, and provides an actionable suggestion and reference for implementersand supplementers and revisionists of standards.

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The content level and particle size distribution of 210Po in
Taiyuan environmental aerosol

Pengxiang Li*, Xiaona Ren, Xuyuan Ma, Feng Zhang, Ruijun Wang, Kaite Lu
China Institute for Radiation Protection，Taiyuan，030006，China1
Email: lipengxiang0214@163.com
Keywords: 210Po，environmental aerosol，content level ，particle size
distribution
In this paper, the isotope tracing method and self-deposition source
preparationis were used to separate the 210Po from the aerosol sample matrix,
then measured the accurate content of 210Po by alpha spectrometer. According
to this experimental procedure, 210Po in environmental aerosols in Taiyuan City,
China was continuously monitored, and one aerosol sample was collected every
month, and a total of 18 months of aerosol samples were analyzed. The results
show that the 210Po content in Taiyuan's environmental aerosols ranges from
0.15mBq/m3 to 1.56mBq/m3, with the average value is 0.82mBq/m3, The
content of 210Po basically showed the law of low in summer and high in winter,with lower levels from May to September and higher levels from October to Marchof the following year, The 210Po content was 0.20~0.67mBq/m3 from May to
September. In order to further understand the particle size distribution of 210Po
in aerosols, the particle size collection and determination of 210Po in aerosol
was carried out by impact-type grading sampler, which was designed to simulate
the human respiratory system, with a total of 8 levels, and the particle size capture
range was 0.4 μm to 10 μm. The experimental results showed that the content of
210Po was the highest in the particle size range of 0.4-0.7μm and 0.7-1.1μm,
and 210Po in aerosols is mainly adsorbed on small-sized particles, and thecontent of 210Po increases with the decrease of aerosol particle size, which may
be related to the specific surface area of aerosol particles. Understanding theadsorption of 210Po on particles of different sizes in the atmosphere provides
basic data for accurate estimation of internal radiation dose.

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O4 - 06#4 Environmental Radiation Measurement and Assessment

Study on the Influence of Large Natural Ventilation CoolingTowers on Atmospheric Diffusion at Offshore NPPs

Junfang ZHANG1,*
Department of Nuclear Environmental Science, China Institute for Radiation
Protection, China1
Email: zhangjunfang@cirp.org.cn
Keywords: Natural ventilation cooling tower, Wind tunnel experiment, Atmospheric diffusion, Numerical simulation, Entrainment effect
The offshore nuclear power plant site adopts a secondary circulation coolingwater supply method, equipped with large countercurrent seawater natural ventilation cooling towers. Due to the impact of cooling towers as tall buildings andthe mist plume emitted from cooling towers on the diffusion of airborne radioactive substances, it is necessary to conduct research on the impact of large naturalventilation cooling towers on the diffusion of nuclear power plant sites based on
the engineering layout. This study summarizes the effects of cooling towers andtheir plumes on the diffusion of airborne radioactive substances through windtunnel experiments and numerical simulations.
According to the characteristics of wind distribution on representative nuclear
power plant sites, in combination with the distribution of residential areas, thedominant wind direction E, the secondary dominant wind direction N, and thewind direction W from the cooling tower to the nuclear island were used as theexperimental wind directions. Consider the influence of cooling tower air inlet
during the experiment.
The site is comprehensively affected by buildings and mountains, and the impact
of buildings such as cooling towers on the site is dominant. The height of theformed building wake area is about 0.8~1.4 times the height of the upstreambuilding, and the length of the wake area is about 10.8~16 times the height of
the upstream building. When the wind direction is approximately 90° to the main
body of the building group, the impact of the cooling tower on the plant site isdominant.
Regardless of whether heat plumes are emitted or not, large natural ventilation
cooling towers have a entrainment effect. The wind tunnel experiment results
show that when the chimney is discharged at a height of 75 meters, the entrainment fraction of the large natural ventilation cooling tower is 29.8% when it isworking, and 1.2% when it is not working. The numerical simulation results show172

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#4 Environmental Radiation Measurement and Assessmentthat when the chimney is discharged at a height of 75 meters, the entrainment
fraction of the large natural ventilation cooling tower is 25.28% when it is working,and 0.75% when it is not working.
Reference
Michioka T，Sato A．Wind tunnel experiment for predicting a visible plume region
from a wet cooling tower [J]. Wind Engineering and Industrial Aerodynamics，2007，95( 8) : 741 － 754． LUCAS M, MART1NEZ P J. On the influence of
psychometric ambient conditions on cooling tower drift deposition [J]. Heat andMass Transfer, 2010, 53B594- 604.

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O4 - 07#4 Environmental Radiation Measurement and Assessment

Concentration behavior and radiation dose of 137Cs in zebrafish
Shaofei CAO, Qingliang FENG, Wenjing GONG
Institute of Nuclear Environmental Science, China Institute for
Radiation Protection/China1
Keywords: zebrafish, concentration ratio, Cs-137, dose rate
The premise for assessing the radiation impact of non-human species is to understand the transfer behavior of radionuclides in organisms. The radiation impact assessment of aquatic organisms firstly requires obtaining the activity concentrations of radionuclides in environmental media (water bodies, sediments,etc.) by means of environmental monitoring or numerical simulation. The nextchallenge is to determine the uptake and transfer parameters of radionuclides byaquatic organisms.
Taking a model organism, the zebrafish （Danio rerio）, as the research object,
an experimental study was conducted on the concentration and excretion of
137Cs (with a half-life of 30.17 years, a representative radionuclide in the liquid
effluents of nuclear facilities) in zebrafish. The activity concentration of 137Cs inwater is 1000 Bq/L. Collected biological samples periodically (day 1, 2, 5, 10, 20,
30, 40, 50, 51), weighed fresh weight, and determined the activity concentrationsof 137Cs in zebrafish using in vivo measurement technology. The activity concentrations of 137Cs in water was measured regularly , and appropriate amountof 137Cs standard solution was added when necessary to maintain its activity
concentration at 1000 Bq/L. After the last batch of sampling and measurement
for the concentration experiment was completed, zebrafish were immediatelytransferred to clean water without adding 137Cs. Collected biological samples
periodically (day 1, 2, 5, 6, 9, 20, 25, 30, 35, 40, and 60), weighed fresh weight,
and the activity concentration of 137Cs in zebrafish was determined by in vivo
measurement technology.
The results showed that the activity concentration of 137Cs in zebrafish increased
linearly with the extension of concentration time. On the 51st day, the activity concentration of 137Cs in zebrafish reached its maximum (5848.7 Bq/kg), and the calculated equilibrium concentration ratio was 5.85. A zebrafish model with multiple internal organs was established in this study to investigate the mechanism of radiation
damage effect in order to protect non-human species. Based on Monte Carlo simulation,the radiation dose rate of Zebrafish was 3.38×10-2μGy/h.
The future research is aimed at comprehensively investigating the relationshipbetween zebrafish radiation dose and biological effects, so as to better evaluate
the radiation effects of radionuclides in liquid effluents on non-human species.

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Preliminary Study on Separation and Concentration for Large
Volume of Krypton by Improved Gas ChromatographXUYUAN MA1,*, JING ZHANG, RUIJUN WANG, PENGXIANG LI, ZHOU LI
China Institute for Radiation Protection, China1
Email: MXY_MONICA@163.COM
Keywords: Improved Gas Chromatograph, Large Volume of Krypton, Separation
and Concentration
When Gaseous effluents from nuclear power plants in China are measured by
gamma spectrometer in the laboratory ,there is a high detection limit about
103~105Bq /m3，the volume of the sample is small and the time of measurement
is short is the reason of high detection limit. Currently, there is no maturesampling method for concentrated gas。
In this experiment, the Gas Chromatograph used in chromatographic separation
method was improved to realize the combination of adsorption method andchromatographic separation method, so as to achieve the separation andconcentration of krypton which was in a large volume of sample. The gas inlet
and the tail gas outlet of the Gas Chromatograph have improved to achieve twostage separation and concentration, then collect Krypton. The volume of Krypton
collected can be quantitatively determined by the calibration curve of Krypton.
Activated carbon was used as adsorption material, MS-13X molecular sieve wasused as column and TCD was used as detector to realize two-stage separationof krypton.
A large volume of sample gas is adsorbed by the concentration column at low
temperature, the impurity gases are purged by the carrier gas after heating thecolumn, and then desorbing the column at high temperature to complete thefirst-stage separation and concentration. The desorbed gas is separated again
by the chromatographic column and Krypton is collected after TCD detector
analyses.
The entire process recovery rate is between 85% and 113%. When the content
of Krypton of the sample gas is 0.1%, the multiple of concentration is about 5
times when the injection volume is about 500 mL, and the multiple ofconcentration is about 12 times when the injection volume is about 1000 mL.

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O4 - 09#4 Environmental Radiation Measurement and Assessment

Analysis of Interference between Tritium and Carbon -14
Measurement

Li BAO, Bing LIAN, Youkun YANG, Chen GUO, Hailan YANG, Xuyuan MA
Key Laboratory of Radiation Environment & Health of the Ministry of Ecology
and Environment, China Institute for radiation proctection /China1
Email:
Keywords: Tritium,Carbon-14,LSC,interference
Tritium and carbon-14, as important radionuclides in gaseous emissions from
nuclear facilities, have attracted much attention because of their large emissions
and large collective dose contribution. For the monitoring of tritium and carbon14 in airborne effluent, cumulative sampling and laboratory analysis are often
used at present. The HTO is captured by water bubbling method (the non-HTO
part in the gas is catalytically oxidized into HTO and then captured), and the CO2
is captured by NaOH solution using bubbling method (the non-CO2 part in thegas is catalytically oxidized into CO2 and then captured). The tritium content in
the water capture solution and the carbon -14 content in the alkali solution aredetermined by liquid scintillation counting, and the concentrations of tritium andcarbon-14 are calculated.In the sampling stage, CO2 will be dissolved in thewater capture solution of tritium, and tritium will also be exchanged and absorbed
in the absorption NaOH solution of carbon-14. The mutual interference between
them in the sampling stage can not be eliminated in the sample preparation stage,
which will cause mutual interference in the subsequent liquid scintillation
measurement, thus affecting the accuracy of the measurement results. In thispaper, the mutual interference of samples containing tritium and carbon-14 inmeasurement is quantitatively analyzed, which provides reference for accurate
analysis of tritium and carbon-14.
The background samples, calibration samples and double-labeled samples oftritium and carbon-14 were prepared, and measured by LSA2000A low-levelliquid scintillation spectrometer. For the measurement of samples both containing
tritium and carbon-14, the dividing line of tritium and carbon-14 counting
window is selected at the maximum energy of low-energy tritium, according to
the different β particle energies emitted by tritium and carbon-14 (βmax, H3=18.6 keV, βmax, C-14=156 keV).The measurement results of samples
containing tritium and carbon -14 in the tritium sample system were comparedand calculated.

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It was found that the counts of double labled sample in the tritium window (50320) was equivalent to the counts of the contribution of single tritium source, thatis, the contribution of carbon-14 counts was not detected in the tritium andcarbon -14 double-labeled samples in the tritium counting mode. For the
carbon-14 sample system, the contribution of tritium can be clearly seen in the
spectrum of samples containing tritium and carbon-14. According to the resultsof tritium counting efficiency in tritium window, carbon counting efficiency intritium window and carbon window, tritium contributes 9.8% to the counting rateof single carbon -14 counting window 50-550 when the content of tritium andcarbon-14 is equal in the sample.
In the measurement of samples both containing tritium and carbon-14, different
types of liquid scintillation spectrometer will give different measurement spectrum
analysis results because of different principles and designs. When LSA2000A
low-level liquid scintillation counter is used as counting instrument, the counting
contribution of carbon-14 can be ignored in tritium counting mode, the counting
interference of tritium on the low-energy part of carbon-14 in carbon-14
counting mode should pay particular attention. When the content of tritium andcarbon-14 in the sample is equal, tritium contributes 9.8% to the counting rateof carbon-14 single-labeled sample. In order to avoid the contribution of tritiumin the measurement of carbon-14, the counting interval of carbon-14 should bedetermined according to the results of tritium and carbon -14 double-labeled
samples.

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O4 - 10#4 Environmental Radiation Measurement and Assessment

Study on Optimal Allocation Method of Dose Constraint Value in
Nuclear Facilities

Hailong CHEN1, Bing LIAN1,*
Key Laboratory of Radiation Environment & Health of the Ministry of Ecology
and Environment, China Institute for Radiation Protection1
Email: lianbing00@sina.com
Keywords: Dose Constraint Value; Optimal; Nuclear Facilities
At present, the dose constraint values of some nuclear bases limit thedevelopment in our country seriously. The reason is that some nuclear bases
have small emissions but occupy large dose constraint values. Some nuclear
facilities are in decommissioned or to be decommissioned, and occupy dose
constraint values.
Therefore, based on the dose constraint value optimization theory in International
Commission on Radiological Protection (ICRP) Report No. 60 and No. 103 andthe International Atomic Energy Agency（IAEA） Safety Standards Series
No.GSG-9 . This paper adopts the statistical analysis method of probability
theory data, the optimization process and method of dose constraint value fornuclear facilities were established, which included irradiation path analysis,
effluent emission optimization analysis and dose constraint value determination.
By using this process and method, the discharge of effluent is statisticallyanalyzed from existing nuclear facilities . It is found that most of the dischargeof effluent obey the normal distribution, while a few facilities do not obey thenormal distribution due to production factors. The potential fluctuation range ofemission and standard variance are used to obtain the wide capacity ofoperational flexibility of two types of nuclear facilities respectively.At the sametime, the annual effective dose levels of representative individuals caused bydifferent nuclear facilities were determined according to the surrounding
environment of nuclear facilities, including the height, inner diameter and flue gas
flow velocity of the air-borne efflux discharge facility, combined with its
meteorological data, population distribution and diet.And the dose constraint
value control values of different nuclear facilities were determined respectively.
The distribution of dose constraint values was optimized to ensure that thereserved dose constraint value of nuclear facilities reached 20%~30%, which
promoted the sustainable and safe development of nuclear facilities.It helps our
nuclear industry to flourish under the premise of safety.

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The Graded Approach in the Radiation Environment Managementof NORM in China

Wang Yan1, Lian Bing1,*, Gu Zhijie1
China Institute for Radiation Protection , Tai Yuan, China1
Email: lianbing@cirp.org.cn
Keywords: NORM, Graded Approach, Radiation Environment Management
The management of the radiation environment of Naturally Occurring Radioactive
Materials (NORM) requires consideration of all stages of its life cycle, including
siting, construction, operation, and decommissioning. The selection andapplication of regulatory control should be commensurate with the relevanthazards and risks. Through the Second National Pollution Source Survey, there
are 464 NORM facilities in China, and it is necessary to carry out supervision andcontrol commensurate with the relevant hazards and risks.
In terms of regulatory responsibilities, the Ministry of Ecology and Environment
supervises and manages the prevention and control of radioactive pollution
nationwide, and the environmental protection departments at or above the county
level supervise and inspect the prevention and control of radioactive pollution in
NORM within their administrative areas. The Directory of Radiation Environment
Supervision for the Development and Utilization of Mineral Resources specifies
the NORM categories and industrial activities that require radiation environment
supervision. It also provides guidelines for the graded approach of environmental
impact assessment and completion environmental protection acceptancemanagement.
The Management Measures for Radiation Environment Monitoring and Information
Disclosure of Facilities for NORM facilities provides the graded approach for theradiation environment monitoring and management of NORM facilities. Provincial
ecological and environmental department establishes a directory of radiation
environment monitoring, while NORM facilities prepare monitoring plans andexecute them. Each provincial-level ecological and environmental department
shall be responsible for summarizing the radiation environment monitoring
information of facilities within its administrative area and submitting it to the
Ministry of Ecology and Environment. The ecological and environmental
departments at or above the municipal level shall be responsible for supervisingand inspecting the radiation environment monitoring of NORM facilities.

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O4 - 11For the decommissioning of NORM facilities, this paper discusses the gradedapproach of environmental impact assessment management based on the resultsof the source survey, pollution types and the relevant hazards and risks of the
proposed decommissioning facilities.
This paper introduces the graded approach of NORM in China in terms ofregulatory responsibilities, environmental impact assessment of construction
projects, and monitoring, and discusses the graded approach of environmental
impact assessment management of NORM facilities decommissioning. A figureof stepwise and graded approach to the regulatory control of NORM in China also
provided.

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Models and Parameter studies for Retrospective Evaluation of
Nuclear Facilities

Jing Kang1, Bing LIAN1,*, Feifei WU1China Institute for Radiation Protection, Key Laboratory of Radiation
Environment & Health of the Ministry of Ecology and Environment, Taiyuan,
030006, China1
Email: lianbing00@sina.com
Keywords: Nuclear sites, Retrospective evaluation, Individual dose estimation
model, Parameter localization
China's large-scale nuclear bases have been in operation since the sixties andseventies of the last century, and have experienced the construction andexpansion of various reactor types and the construction of nuclear fuel cycle
facilities. After decades of smooth operation, multiple reactors or multiple
facilities at one site have formed in the nuclear large-scale base. With theconstruction of nuclear facilities and the continuous changes in the surrounding
environmental conditions, it is necessary to carry out a retrospective assessmentof the nuclear base as a whole as soon as possible, with the aim of accurately
grasping the emission situation of the nuclear base in recent years and the actual
impact on the surrounding environment. Evaluation of the dose of staff in the
nuclear base, surrounding public groups and other relevant personnel andaccurately grasp the dose of the population, can provide data support for thesustainable development of the nuclear base.
The retrospective assessment of nuclear sites mainly includes two aspects: thefield distribution of radionuclide concentrations within the nuclear site and in thenear-field area, and the dose assessment of staff and the surrounding typical
population. In the accounting emission sources of nuclear sites, this study
focuses on the differences in radionuclides emitted by different types of facilities,
the actual emissions of different operating conditions of facilities, and the validityof emission data. In the estimation and evaluation of concentration field
distribution, this study focuses on the complex underlying surface of the structure
and the sedimentation regularity of typical radionuclides. In terms of dose
assessment, this study focuses on the dose assessment models of several
representative populations, the authenticity of life trajectories and the localizationof parameters, especially the biological transfer parameters affected by foodroutes.

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O4 - 12In this study, a retrospective evaluation method and personal dose estimation
model of nuclear bases were constructed, and the dose estimation results oftypical populations were compared with the estimation results of other foreign
nuclear bases. Conclusion that China's nuclear bases have been operating wellfor many years and have not yet had a significant impact on the surrounding
environment is credible.

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Application of Deconvolution Method to Alpha-ray Energy
Spectrum to Improve Performance of Radiation Airborne
Particulate Monitor

Makoto Sasano1,*, Masateru Hayashi1, Yusuke Yanagawa4, Masatoshi Kai4,
Yasushi Nakano4, Yoshitsugu Osawa4, Shunsuke Kurosawa2, Yuki Morishita3
Mitsubishi Electric Corp./Japan1, Tohoku University/Japan2, Japan Atomic
Energy Agency/Japan3, Mitsubishi Electric Plant Engineering Corp./Japan4
Email: Sasano.Makoto@db.MitsubishiElectric.co.jp
Keywords: Airborne particulate monitor, alpha-ray energy spectrum,
deconvolution
At nuclear facilities and decommissioning sites, monitoring radioactivity
concentration in airborne particulate is necessary to prevent worker exposure. To
avoid internal exposure, we should detect alpha-decay radionuclides. When wemonitor alpha-decay radionuclides in airborne dust, we want to measure theconcentration of only artificial radionuclides. However, there are naturalradionuclides in the air. Therefore, it is necessary to separate artificial
radionuclides and natural radionuclides in the measurement. Therefore, in thisstudy, we analyzed radionuclides in the collected dust and investigated a method
to measure the radioactivity concentration separately for artificial and naturalradionuclides.
There are several methods to analyze radionuclides. Among them, the analysis
by energy spectrum is widely adopted. When this method is applied to the energy
spectrum of alpha rays, energy loss between the measurement target and thedetector becomes an issue. This is because the original alpha-ray energy
information is lost due to the energy loss, and therefore, we are not able toidentify the nuclide from the detected energy. Furthermore, in the energy
spectrum, the alpha-ray peaks have a tail structure overlapping each other. So,
it is not possible to separate the nuclide as it is.
This study applies the deconvolution method to the alpha-ray energy spectrum
to separate artificial and natural radionuclides. We need to prepare response
functions to obtain the deconvolved energy spectrum. Therefore, we createdresponse functions for various alpha-ray energies with a radiation simulation
toolkit. Utilizing these response functions, we deconvolved the alpha-ray energy
spectra measured while collecting dust on filter paper. To make the response

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O4 - 13function accurate, we prepared a model including the distance between themeasured object and the detector and the structure of the light shield in detail.
We acquired energy spectra every 10 minutes in our laboratory and deconvolved
them. The deconvolved spectra had three clear peaks at 6.0, 7.7, and 8.8 MeV.
These energies were consistent with those of 212Bi, 214Po, and 212Po. The
deconvoluted energy spectra showed that only a few measurements were
included in the energy range due to artificial radionuclides, 4.0 to 5.8 MeV. Fromthe measurements, the decision threshold for artificial radionuclides was lessthan 1E-7 Bq/cm3 with the 10 minutes measurement. These results show that it
is possible to measure artificial and natural radionuclides separately with thedeconvolved alpha-ray energy spectra. And it was demonstrated that artificial
radionuclides could be monitored down to low radioactivity concentrations.
Acknowledgement
This work was supported by JAEA Nuclear Energy S&T and Human Resource
Development Project through concentrating wisdom Grant Number
JPJA18P18071964.

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#4 Environmental Radiation Measurement and Assessment#4 Environmental Radiation Measurement and Assessment

CFD Simulation of Resuspension Release of Radioactive Aerosol
Particles from Contaminated Sites

Jia-chen CHEN*, Hai-long CHEN, Bing LIAN , Yan WANG
China Institute for Radiation Protection,China1
Email: 18903505148@163.com
Keywords: suspended particles,CFD method,airflow field
Once radioactive substances cause radioactive contamination of soil in a certain
area through atmospheric deposition, surface water, and groundwater, theparticles in the soil may be resuspended from the ground to the air due to physicaldisturbances caused by wind or human activities. For nuclides such as uraniumthat are difficult to transfer to humans through the food chain, internal exposure
caused by inhalation of resuspended particles is a route that must be taken
seriously. Therefore, understanding the resuspension behavior of particles is ofgreat significance for the study of control measures for suspended particles.
Based on the CFD method, this paper simulates the resuspension process andtrajectory of radioactive aerosol particles caused by the air inlet, and analyzes
the influence of the air inlet and outlet settings on the distribution characteristicsof the airflow field in the resuspension box.
The results show that when the wind speed at the air inlet is 10m/s, the airflow
in the resuspension box is evenly distributed, and the vortex flow is enough tocause the resuspension of radioactive aerosol particles on the gas-solidinterface, and the amount of resuspension of particles is large, which isconvenient for aerosol sampling and measure.
This simulation calculation can provide a basis for the subsequent experimental
design of the settlement and resuspension of radioactive aerosol particles in thenatural environment and disturbed environment, and provide guidance for theradiation protection actions of people working or active in radioactively
contaminated areas.

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O4 - 15#4 Environmental Radiation Measurement and Assessment

Parameter sensitivity analysis of different ecological environments
for a tritium environmental transfer model

Junyu CHENG1, Rentai YAO1,*
China Institute for Radiation Protection1
Email: yaorentai@sina.com
Keywords: Tritium transfer, compartmental model,sensitivity analysis
The sensitivity analysis of key radio-ecological parameters attracted extensive
attention of the radionuclide transfer modeling in the environment. In this paper,a dynamic compartmental model of tritium environmental transfer has been
established, and the model validation was conducted through model intercomparison on the basis of three meteorological conditions of the Hypothetical
Scenario from IAEA EMRAS international programme. Through selecting three
nuclear power plant sites as the representative different radio-ecological zones
in China, i.e. Daya Bay NPP, Tianwan NPP and Hongyanhe NPP, the sensitivityof model to the different environmental parameters in agricultural and terrestrial
ecosystems was studied by simulating the distribution of tritium concentration indifferent environmental compartments around nuclear power plants after tritium
short-term releases for the different hypothetical scenario under real radioecological conditions. The results show that the sensitive parameters in themodel mainly focus on meteorological conditions, local soil properties andevaporation, the parameterized method of plant growth process and so on, andthe model can be used to simulate the environmental transfer of tritium shortterm release in the terrestrial ecosystem reasonably.

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O5 - 01#5 Radiological Risk Management

Radiological Assessment of Loss of Coolant Accident from Shin
Kori Nuclear Power Plant

Moses Oboo, Jouyoul Kim*
Kepco International Nuclear Graduate School (KINGS), Department of Nuclear
Power Plant Engineering, South Korea1
Email: jykim@kings.ac.kr
Keywords: LOCA, RASCAL code, GEN II code, Exposure Pathways, Individual
effective doses
This paper aims to evaluate the impact of the radionuclide releases from thesevere loss of coolant accident (LOCA) of the APR1400-Shin Kori nuclear power
plant, the latest advanced Pressurized water reactor (PWR) type developed by
the Republic of Korea. In a PWR, water is used as both the coolant and themoderator for the reactor core, where it absorbs heat from the fuel rods before
circulating to the steam generator.
A LOCA occurs when there is a loss of coolant from the RCS, which can becaused by a breach in the piping or a failure of a component such as a pump or
valve. A LOCA can result in a rapid decrease in the water level in the core, which
can cause the fuel rods to overheat and potentially melt. To prevent or mitigate
the effects of a LOCA, PWRs are designed with a number of safety features,
including emergency core cooling systems (ECCS) that rapidly inject coolant into
the core after a LOCA, and containment structures that prevent the release ofradioactive materials to the environment in the event of a breach in the primarycontainment. However, in the case of a severe accident, all these mitigation
measures usually fail, and radioactive materials are released to the environment,
posing significant health risks to workers and members of the public (S. N. A.
Sulaiman, etal 2018).
The LOCA will be simulated using the RASCAL code to obtain relevant source
terms (J. V Ramsdell, etal; 2012). This study will consider the four radionuclidesof 133Xe, 131I, 134Cs, and 137Cs which pose serious risks to human health andthe environment (S. Hashimoto, etal; 2022) and this will then be input into the
GEN II code to calculate individual effective doses in 16 different compass directions at appropriate distances from the NPP site. The study will consider several parameters that can influence individual effective dose, including the magnitude of the radionuclide releases, the consumption rate of common foods bypeople living near Kori NPP, the exposure pathways considered, indoor and out-187

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O5 - 01

door inhalation rates and the meteorological data of the area.
The output results will provide the contributions of each exposure pathways toradiation exposure, the individual effective doses at specific directions and distance from the NPP and depending on the magnitude of the doses, it will provide
necessary information for emergency responders on the particular area where
people will need either sheltering or evacuation in case of such a severe accident
on the Shin Kori Nuclear Power Plant.
Reference[1] J. V Ramsdell, G. F. Athey, and J. P. Rishel, RASCAL 4: Description of models
and methods. United States Nuclear Regulatory Commission, Office of Nuclear
Security and Incient Response, 2012. [2] S. N. A. Sulaiman, F. Mohamed, andA. N. Ab Rahim, Radioactive release during nuclear accidents in Chernobyl and
Fukushima, IOP Conference Series: Materials Science and Engineering, vol. 298,
no. 1, p. 12011, 2018. [3]S. Hashimoto, M. Komatsu, and S. Miura, Radioactive
Materials Released by the Fukushima Nuclear Accident, Forest Radioecology inFukushima, Springer, pp. 1-10, 2022

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O5 - 02#5 Radiological Risk Management

Radiological Risk Assessment of Safety Exit Sign Containing
Tritium.

Yeijin BANG1, Jimin SHIN1, Sunlee SHIN1, Hee SEO1,*, Byeonghyeon PARK2,
Sangmin LEE2
Department of Applied Plasma and Quantum Beam Engineering, Jeonbuk
National University, Korea1, Department of Radiaion Regulation, Korea Institueof Nuclear Safety, Korea2
Email: hseo@jbnu.ac.kr
Keywords: Exposure scenario, Radiological risk assessment, Safety exit sign,
tritium, Regulation exemption
In the Republic of Korea, some consumer products containing radioactive
isotopes are excluded from regulation through the "Regulation on Substances
Excluded from Radioactive Isotopes", in addition to the general regulation
exemption criteria recommended by the IAEA. Based on this, consumer products
are allowed to be used and disposed of by the public without particular restraint.
One of such products is the safety exit sign, which contains up to 925 GBq oftritium, exceeding the general regulatory exemption criteria of 1 GBq. Safety exit
signs are essential in guiding individuals during power outages and emergencies,
such as fires. Recently, there has been a trend towards replacing them with LEDs;
however, due to the relatively long half-life of tritium (i.e., 12.3 years), whichallows for long-term use after a single installation, tritium-based safety exit signs
are still widely used. Therefore, in this study, we aimed to develop predictabledisposal and accident exposure scenarios and the radiological risk was evaluatedto examine the validity of the regulatory exemption of tritium-containing safety
exit signs.
The gaseous form of tritium inside these signs is absorbed by the phosphor in
the borosilicate glass tube to emit light. As tritium emits low-energy beta particles
with short ranges (i.e., about 6 μm in water), only internal exposure wasconsidered except for external exposure. The exposure scenarios considered
included disposal at a recycling center, breakage during replacement by anindividual, breakage during transportation to a recycling center, and a fire in ahigh-rise building. Exposure scenario factors were developed by considering therecycling rate, distribution volume, and others according to the domestic
situation. For each scenario, dose assessments were conducted for the generalpublic, firefighters, and workers.

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O5 - 02

The results of dose assessments for all scenarios were less than 10 µSv/y for
general scenarios (i.e., disposal) and 1 mSv/y for low probability scenarios (i.e.,
accident). Although the scenarios were developed conservatively, all met the
IAEA regulatory exemption criteria.
In this study, we developed the disposal and accident exposure scenarios andevaluated the radiological risk of safety exit signs containing tritium above thegeneral regulatory exemption criteria of 1 GBq. It was found that the radiation
doses for all exposure scenarios were below the exemption levels specified by
the IAEA regulations. Although the safety exit sign poses no significant
radiological risk, a safety management guide will be developed to prevent
indiscriminate disposal and to inform proper handling procedures in the event ofan accident.

190

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O5 - 03#5 Radiological Risk Management

Radiation Dose Assessment of Thorium-containing Gas Mantle
for Predictable Exposure Scenarios

Jimin SHIN1, Yeijin BANG1, Sunlee SHIN1, Hee SEO1,*, Byeonghyeon PARK2,
Sangmin LEE2
Department of Applied Plasma and Quantum Beam Engineering, Jeonbuk
National University, Korea1, Department of Radiation Regulation, Korea Instituteof Nuclear Safety, Korea2
Email: hseo@jbnu.ac.kr
Keywords: Dose assessment, Exposure scenario, Gas mantle, MicroShield SW,
MCNP
Each country established specific exemption criteria for consumer products
containing radioactive isotopes in amounts greater than the general exemption
criteria based on the recommendations of the IAEA. One of the major conditions
set by the IAEA is that the result of the dose assessment for predictable exposure
scenarios by using the product should be below certain limits. However, since
product specification and distribution amount are changed over time, a periodic
review of validation of the specific exemption criteria based on the dose
assessment is necessary. In this study, we assumed exposure scenarios basedon the cycle of use of thorium-containing gas mantles that correspond to thedomestic specific exemption criteria and evaluated the exposure dose for eachscenario. In addition, the reliability of the MicroShield software used for external
dose assessment was examined by comparing them with the results of the Monte
Carlo simulation using the MCNP code.
The gas mantles that meet the specific exemption criteria are entirely imported,and the distribution amount was assumed based on the amount imported into
South Korea in 2021. The thorium content in the mantles was assumed to be themaximum amount allowed under the regulatory exemption standards of 700 Bq/g,
and the radiation doses were evaluated considering the exposure pathways foreach daughter nuclide of thorium. Exposure scenarios were transport, distribution,
usage, accident, and disposal. It was assumed that the mantles imported into
South Korea by ship were transported to logistics centers and retail stores by 40foot cargo trucks and one-ton delivery vehicles, respectively. Workers handlingthe mantles at the logistics centers and retail stores were considered in the
distribution scenarios, and the usage scenarios were assumed for camping.

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Accident scenarios included traffic accidents, fires, misuse, and ingestion bychildren while disposal scenarios were landfill and incineration, considering thedomestic general waste disposal methods. External and internal exposure doses
were evaluated for each scenario. The reliability of the MicroShield software used
for external dose assessment was examined by comparing it with the MCNP
simulation results for two cases (i.e., transport and accident).
As a result, the doses for the general scenarios of transport, distribution, usage,
and disposal, as well as the low probability accident scenarios were all below the
IAEA regulatory exemption standards (i.e., less than 10 μSv/y for the generalscenarios and 1 mSv/y for the low probability scenarios). In addition, whencomparing the results of the MicroShield SW and MCNP calculations for theaccident scenario without shielding and for the transport scenario with steel
shielding, there were large differences of up to 200%.
In this study, we developed exposure scenarios and evaluated the dose for eachscenario using thorium-containing gas mantles. Additionally, the reliability of theMicroShield SW used for dose calculations was examined. The dose assessment
results showed that, even with conservative assumptions, the radiation doses
from all predictable scenarios met the IAEA criteria. Significant differences wereobserved between Microshield SW and MCNP calculation results, which is
expected to be due to the influence of the build-up factor (BUF) used in
MicroShield calculations.

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O5 - 04#5 Radiological Risk Management

Assessment of Source Terms and Potential Doses Due to SteamGenerator Tube Rupture of VVER-1200 at El Dabaa Nuclear
Power Plant in Egypt

waad saleh1, Juyoul kim1,*
department of radiation safety and environmental protection, Kepco nuclear
graduate school, south korea1
Email: jykim@kings.ac.kr
Keywords: Radiological Assessment, SGTR, VEER120
Four VVER-1200 reactors are planned for the El Dabaa nuclear power generationsite in order to incorporate nuclear generation into the energy mix. The
unanticipated occurrences can happen anywhere in the world at any time. For
example, recently, Turkey and Syria witnessed severe earthquake during themonth of February 2023. Egypt has also experienced 10 earthquakes in this year
till today though less severe but in future the intensity of earthquakes can bestronger. The events like this pose a strong threat to the nuclear power plants,e.g., the Fukushima Daiichi accident. It is imperative to ascertain the importanceof assessing NPP safety measures. The El Dabaa site may also need to go under
several post-accident safety assessments before operation.
The release of Radionuclides, following a major accident, may pose threat to thehealth of nearby population. The life-threatening risk of nuclear radiation,
demands an assessment of the possible radiological hazards, following ahypothetical accident against which plant’s safety systems have been
compromised. Therefore, analysis of the consequence of such an accident must
be taken into consideration while deciding what precautionary measures to betaken. This paper calculates source term and potential doses in case of SteamGenerator Tube Rupture (SGTR) Accident.
During the period of 10 years (2013-2023), the wind rose for meteorological data
for the winter season has been created using the Wind and Rain Rose Plots
software (WRPLOT) to specify the most affected area due to the direction of theplume, the wind speed, and the stability class. Moreover, Radiation Assessment
System for Consequence Analysis (RASCAL) Code is used for the simulation of
SGTR accident to calculate the projected doses and source term radioactive
nuclides. We also used Minitab software to represent the outcomes of ourstatistical investigation for each meteorological data, calculated does and source

193

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O5 - 04

term radioactive nuclides.
The calculated Maximum Total Effective Dose Equivalent (TEDE) is 2.9 mSv at
0.16 km from the nuclear power plant site and it reaches 0.12 mSv at the distanceof 3 km. The thyroid does at 0.16 km is 0.25 mSv. However, the max distancethat it can travel is 1.6 km with the dose level of 0.015 mSv at 1.6 km. The source
term maximum dose stands at 1.2×1011 Bq and 3.4×109 Bq for 137Cs and 131I
nuclides, respectively.
Acknowledgement
This research was supported by the 2023 Research Fund of the KEPCO
International Nuclear Graduate School (KINGS), the Republic of Korea.

194

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O5 - 05#5 Radiological Risk Management

Thermal Safety Assessment of Transport Containers forRadioactive Materials

Changwu Wang1,*, Yuhang Zhang1, Yiren Lian1, Lei Chen1, Zhipeng Wang1, Qian
Sun1
China Institute for Radiation Protection/China1
Email: wangchangwu@mail.nwpu.edu.cn
Keywords: Thermal safety, Heat resistance test, Transport safety
With the rapid development of nuclear energy, the number of packages fortransporting radioactive materials has increased dramatically, and the safety oftransporting radioactive materials has attracted more attention. Transport
containers for radioactive materials are physical safety barriers for thetransportation of radioactive materials. By evaluating the safety performance oftransport containers for radioactive materials, they can effectively safeguard
human health, protect the environment, and promote the development andpeaceful use of nuclear energy and nuclear technology. The thermal safety
evaluation of radioactive material transport containers is an important content of
the safety performance evaluation of radioactive material transport containers.
The design of transportation containers can usually be verified by simulation
analysis methods, test methods, or a combination of both. Considering the
economy and reliability of the results, a combination of test and simulation
analysis is often used. This article takes the thermal safety evaluation of AF typetransport containers as an example, and summarizes the experience of thermal
safety evaluation of radioactive material transport containers in combination withdaily work.
The thermal safety evaluation of radioactive material transportation containers isdivided into two parts: simulation analysis and experimental verification. Firstly,
based on the structural characteristics of transportation containers, normaltransportation conditions, and transportation accident conditions, corresponding
physical and mathematical models are established for simulation analysis andcalculation. Then, a heat resistance test was conducted according to Regulations
for the Safe Transport of Radioactive Material (IAEA SSR-6 (Rev. 1) - 2018). The
test results showed that the temperature at the inner cover of the transport
container was less than 104 ℃, which was lower than the requirement of theacceptance criteria that the temperature at the inner cover should not be higher
than 252 ℃, meeting the acceptance criteria. Comparing the simulation

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O5 - 05

calculation results and test data of key components, it was found that thetemperature data obtained from the simulation calculation and the test data werein good agreement, with a maximum temperature difference of less than 14%.
The results show that using simulation analysis and experimental techniques tocollect and analyze temperature data of key components of transportation
containers can effectively evaluate the thermal safety performance of packages.
After summarizing several thermal safety evaluations of radioactive materialtransportation containers, it was found that the accuracy of simulation results
depends on the rationality of the calculation model, the accuracy of boundary
conditions, and the experience of computational personnel. The measurement
method of heat resistance test can also affect the accuracy of test data.
Therefore, simulation analysis should pay attention to the key points of modeling,boundary conditions, and result processing. Heat resistance testing should
enrich measurement methods, reasonably select temperature measurement
points, and further explore test results. Compare the measured values of multiple
parameters with the simulation results to verify the reliability of the simulation
calculation.

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O5 - 06#5 Radiological Risk Management

Discussion of “Application of knowledge management inradioactive waste management”

enbang liu1, xiliang guo1,*, chao gao1, yanjiao wang1, xu han1, yahui xi1,
xiaojun yan1
China Institute For Radiation Protetion1
Email: Guoxl214@126.com
Keywords: Radioactive waste management, Information management systems,
Knowledge management ,Knowledge management systems【Purpose】The generation of radioactive waste has a tendency to rise as thenuclear energy industry develops, and it is becoming increasingly challenging tocontrol the radioactivity's entry into the warehouse. The informational framework
for managing radioactive waste can increase data entry accuracy, lower data
management error rates, ensure the traceability of data pertaining to production,
treatment, storage, transportation, disposal, and other processes, and lower radiation exposure levels for staff.【Method】This paper is based on the analysis
of numerous radioactive waste management cases. By examining the information
management application cases and knowledge management application cases,
it facilitates the development of knowledge management and the innovation andresearch of management techniques for radioactive waste managers.【Results】Managers can intuitively ascertain the degree of radioactivity of waste, thelocation of various waste management sections, and the state of radioactive
waste management through the use of information management systems. Nevertheless, the conventional information management system is unsuitable and nolonger able to meet the needs of managing radioactive wastes since information
on radioactive waste is readily lost and cannot be shared.【Conclusion】By systematically integrating recording of radioactive waste information and sharing information, knowledge management systems can increase the safety of managing
radioactive waste. The idea of a comprehensive system for knowledge management of radioactive waste is advanced, and theoretical support is provided forthe rapid creation of a national information management system for radioactive
wastes using a combination of domestic and international application cases andcutting-edge knowledge management technologies.

197

Abstracts - Oral

O6 - 01#6 Education, Training and Policy in Radiation Safety

Tutorial Development for Digital Twin Simulation in Scenariobased Linear Accelerator Daily QA Training

Wonhyeong Lee1, Joengeun Kim1, Yoonsun Chung1,*
Department of Nuclear Engineering, Hanyang University, Korea1
Email: ychung@hanyang.ac.kr
Keywords: Radiotherapy, Quality Assurance, Training, Digital Twin, Simulation
Radiotherapy requires appropriate training for the safety and accuracy oftreatment, but the most desirable field training has many practical limitations mostof the time. The use of simulation may offer a potential alternative to reducing
burden while maintaining educational effectiveness. And further introduction of
new technologies such as digital twin and virtual reality can enrich the effect of
the simulation by maximizing synchronization with reality.
A previous study established a virtual training environment in a linear accelerator(LINAC) treatment room. We also designed daily quality assurance (QA) content
based on the logic linking training scenarios and simulation objects. However,for those who are new to the field of radiotherapy or who are not familiar withsimulation content, this training method can be confusing and difficult. In thisstudy, we intend to provide higher-quality education by developing additionaltutorial content in the simulation so that beginners can adapt to QA work moreeasily.
The overall simulation was built with the Unity® software (ver. 2021.2.12f1), oneof the representative game engines. As reference models, we chose theTruebeam™ STx of Varian for the LINAC machine and the Daily QA™3 of
SunNuclear for the dosimetry device. AutoDesk Inventor Professional 2019 andFusion 360 were used for precise modeling and objects were exported to theUnity environment. The procedures of LINAC daily QA were organized in the formof scenarios referring to the AAPM Task Group 142 report, and implemented as
logic by linking them with simulation objects.
Before developing the tutorial content, we first separated each QA scenario sothat players could select and perform each QA scenario independently. The
parallel configuration of these scenarios is intended to make the training morestructured and concise for players new to the tutorial.

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#6 Education, Training and Policy in Radiation Safety

Each QA scenario has its own tutorial structure, and players can choose whether
or not to activate the tutorial before running the scenario. An activated tutorial
guides the player through the entire process of the selected QA scenario insequence. Objects or user interfaces (UIs) that require interaction in eachsituation are highlighted in an outline using ‘outline shader’. The outline shader
operates on specific pre-coded elements and points as the tutorial progresses.
Additional explanation required for the tutorial is provided by denoting text scripts
on part of the display. The text scripts not only provide instructions to the player
for the scenario but also serve to compensate for the educational limitations of
the simulation. Text scripts can present in-depth information such as tips thatcannot be implemented in the simulation, and notes/instructions in actual QA
work.
And in addition to tutorials on QA scenarios, we also developed basic tutorials
for first-time players. The basic tutorial introduces players to basic concepts andcontrols, helping beginning players to easily acclimate themselves to thesimulation before primary training.
In this study, we added a tutorial to a previously developed LINAC daily QA
training simulation. Tutorials can improve the accessibility, comprehension, andconcentration of trainees on this simulation-based training method. These
additional functions can enhance the effectiveness of the simulation-based
training method and are expected to contribute to the expansion of training
content for radiotherapy professions in the future.

199

Abstracts - Oral

O6 - 02#6 Education, Training and Policy in Radiation Safety

Report on the Association for Radiation Science and Applications
as a Collaboration of Academia and Industry in Japan

Tetsuo Iguchi1, Hiroyuki Takahashi2,*
Nagoya University1, The University of Tokyo2
Email: leo@n.t.u-tokyo.ac.jp
Keywords: Collaboration, Industry-Academia, Association
We established a research forum called “Radiation Physics Forum” in 2008 withthe purpose of exchanging information on studies which went wrong. This forum
worked as a place for discussions and communications on the applications,
measurement techniques, device and detector developments using ionizing radiation. The forum workshop was held with paying attention to make the separation between the material/device side and detector side less and invited people
from materials/devices manufacturers.
As the growth of the number of participants, this forum has been enhanced andupgraded to become “Industry-Academia Collaboration Committee for Radiation
Science and Applications” as one of the Industry-Academia Collaboration Committees in the Japan Society for Promoting Science (JSPS) from the year 2012 to
2021. This committee had nearly 30 academics and people from 30 industrial
companies. The committee organized 4 regular meetings a year. Each meetinghad 4 or 5 lectures nearly one hour long. The committee was fruitful to develop
industrial-academia collaborations in nuclear, medical, security and basic science fields.
Additional to the regular meetings, a sub-workshop called “Radiation Measurement Forum Fukushima” were held twice a year for sharing information on environmental dose rate and radio-activity measurements for the reconstruction of
Fukushima prefecture from the nuclear power plant accident of Tokyo ElectricPower Company.
One of the purposes of the committee was inviting a large-scale international
conference on radiation measurements in Japan. After some years of campaign,
we succeeded to have IEEE Nuclear Science Symposium and Medical Imaging
Conference in 2021, although it was just a virtual conference due to COVID-19,
we plan to have another conference in Yokohama. Unfortunately, as JSPS has
stopped the support for the secretariat of such industry-academia collaboration,
we decided to continue the same activity and established the Association for
Radiation Science and Applications.

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#6 Education, Training and Policy in Radiation Safety

In this presentation, methods for organizing the association and selecting thelecture topics are shown. We hope similar associations might be organized in
Korea, China and in other countries. We expect more communications between
associations and international collaboration in the future.

201

Abstracts - Oral

O7 - 01#7 Radiological Emergency Planning and Preparedness

Validation and application of migration diffusion mode in nuclear
accident consequence evaluation system

MINGHUA LYU1, RENTAI YAO1,*, JUNFANG ZHANG1, DUOXIN ZHAO1, SHA
HUANG1
Dapartment of Nuclear environmental science research, China Institute for
Radiation Protection, CHINA1
Email: cirpdq@163.com
Keywords: Consequence accessment, validation, multiscale, diffusion
Nuclear emergency is a means to mitigate the consequences of incidents during
the serious accidents of nuclear facilities and nuclear activities, natural disasters
and terrorist events. Nuclear accident consequence assessment technology is anecessary technical means for decision-making during nuclear emergency
response.
The project team has developed a national nuclear accident consequenceassessment and decision support system (NACADOS) with independent
intellectual property rights and suitable for China's national conditions throughresearch on multiscale weather forecasting methods, airborne radioactive
material migration and diffusion simulation technology, and comprehensive
evaluation of system effectiveness. It can meet the needs of radiationconsequences and response of any global nuclear nuclear and radiation
emergencies, and improve the ability of nuclear emergency response.
In view of the complex, rapid, and multiscale characteristics of radioactive
pollutants in the atmospheric migration process, and considering the actual
emission scenario, this article has developed a multiscale meteorologicalforecast simulation function to provide a grid wind field for diffusion by
conducting research on the migration and diffusion simulation technology ofairborne radioactive substances, as well as the applicability and simulation
performance of different modules. At the same time, diffusion modules suitablefor different scales and environmental conditions have been developed, including
trajectory/particle modules suitable for large scales. It meets the requirements ofmulti-source, variable source, and mobile source simulation, and achieves
migration and diffusion simulation at any point and for any time in the world.
Through comparison with on-site monitoring data, high-quality tracer
experimental results (such as ETEX, Urban2000, etc.), and simulation results ofconsequence evaluation systems (JRODOS, RASCAL, etc.), the system's202

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#7 Radiological Emergency Planning and Preparedness

functionality, performance, and effectiveness are tested in three aspects:
physical module validation, system testing, and engineering software testing. The
results show that the system calculation results are true and accurate, and thesystem operates stably and reliably. For example, NACADOS simulation can
obviously reflect various processes in the UNSCEAR report. However, due to thelack of precipitation observation data, the range and magnitude of surface
sediment concentration in the northwest direction are different from thosemonitored. However, in areas with small precipitation impact, the coincidence
factor between the system simulation results and the measured data hassignificantly improved, with the coincidence degree reaching 80% within 10 times,
68% within 5 times, and 58% within 3.5 times, Up to 33% within 2 times.

203

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O7 - 02#7 Radiological Emergency Planning and Preparedness

Investigation on Decision-Making Criteria for Public Protective
Actions in Nuclear Emergencies

Wi-Ho Ha1,*, Chanki Lee1, Ilje Cho1, Won Tae Hwang1, Hyun Ki Kim1
Department of Nuclear Emergency Preparedness, Korea Atomic Energy
Research Institute/Korea1
Email: hwh@kaeri.re.kr
Keywords: nuclear emergency, public protective actions, generic criteria,
projected dose, emergency preparedness and response
At the preparedness stage of nuclear accidents, national protection strategies
should be developed not only to avoid severe deterministic effects but also to
reduce the risk of stochastic effects. For effective implementation of theprotective actions, decision-making criteria should be pre-determined bynational authorities. The International Atomic Energy Agency (IAEA) publishedsafety standards reports on the safety requirements and safety guides dealing
with the decision-making criteria called generic criteria for the public protective
actions in nuclear or radiological emergencies. Therefore, the NSSC (Nuclear
Safety and Security Commission) in Korea has been trying to update the national
decision-making criteria since 2020.
Public protective actions are mainly divided into urgent protective actions andearly protective actions. The urgent protective actions include sheltering,evacuation and ITB (Iodine Thyroid Blocking), whereas the early protective
actions include temporary relocation and restriction of consumption on food, milk
and drinking water. These protective actions should be timely implemented
depending on the response phase as well as radiological impacts at the on- andoff-sites.
To determine the national decision-making criteria, national reference levels
should be assigned in advance. In general, reference levels of 20 to 100 mSv, asa residual dose, are recommended to be applied in emergency exposure
situations. The decision-making criteria as protection quantities (effective doseor equivalent dose) can be provided with a projected dose or received dose
based on the national reference levels. The decision-making criteria adopted in
24 countries were investigated and the national criteria for each protective action
were suggested in this study.

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O7 - 03#7 Radiological Emergency Planning and Preparedness

A visual-based positioning and guidance approach for nuclear
emergency response manipulator

Qi Lv1,*, Xiaoyu Zhao1, Yuyao Tang1, Yi Liu1, Yapeng Yang1
China Institute for Radiation Protection, China1
Email: 346308962@qq.com
Keywords: nuclear emergency response, response disposition manipulator,
machine learning, visual guidance
The emergency response capability for nuclear and radiation accidents is related
to the safe and stable operation of nuclear facilities, and response disposition isa key part of the emergency response system. The manipulator can enter
accident areas that cannot be reached by humans to work, reducing the impact
of radiation on personnel. However, the current operation mode of nuclear
emergency manipulator is generally teleoperation with a low degree of automation,and operators cannot accurately estimate the position relationship between thetarget and the manipulator through 2D images, making it difficult to timely control
the accident. Meanwhile, low disposal efficiency means that the manipulator willwithstand long-term radiation exposure, putting forward higher requirements toradiation-resistant properties.
Aiming at the drawbacks of teleoperation of nuclear emergency manipulator, thispaper establishes a manipulator hand-eye vision system. Due to the great image
recognition performance of machine vision, a two-stage visual positioning andautonomous guidance method is proposed in this paper. First, capture a video
stream from a front camera fixed to a mobile platform, manually initialize theobject to be disposed of (such as a bare radioactive source) using a rectangularframe, transform the target coordinates to the manipulator coordinate systemthrough a hand-eye coordinate matrix, and guide the end tool of the manipulator
to the target area to complete rough positioning. Secondly, capture a video
stream from a camera fixed to the end of the manipulator, use the TLD single
target long-term tracking architecture to learn online and continuously update
the salient feature points of the tracking module and the target model of thedetection module, dynamically locate unknown radioactive materials, andultimately control the robotic arm to approach the target position to complete
accurate guidance.
The performance of the visual algorithm is demonstrated to be superior byexperiments on OTB and VOT benchmarks, and the effectiveness of the guidance

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O7 - 03approach is demonstrated through experiments on a 6-DoF nuclear emergency
manipulator.
To sum up, the new approach can successfully complete nuclear emergency
disposal tasks such as recovery of the radioactive source, greatly improving
disposal efficiency and reducing the risk of radiation accidents.

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#7 Radiological Emergency Planning and Preparedness#7 Radiological Emergency Planning and Preparedness

Machine Learning-based Denoising Technique for Gamma
Emission Tomography to Inspect Partial Defects of Spent Nuclear
Fuel

Hyung-Joo Choi1, Yoon Soo Chung1, Hyun Joon Choi2, Kyunghoon Cho3,
Hakjae Lee3, Hyung Cheol Lee4, Yong Hyun Chung1, Chul Hee Min1,*
Department of Radiation Convergence Engineering, Yonsei University, Republicof Korea1, Department of Radiation Oncology, Wonju College of Medicine,
Republic of Korea2, ARALE Laboratory Co. Ltd, Republic of Korea3, Nuclear
Material Analysis Team, Korea Insitute of Nuclear Nonproliferation and Control,
Daejeon, Republic of Korea4
Email: chmin@yonsei.ac.kr
Keywords: Spent nuclear fuel, Machnie learning, Gamma emission tomogrpahy,
Partial defects, Denoising technique
The spent nuclear fuel (SNF) is byproduct of nuclear power generation, and it
includes fissile materials such as U235 and Pu239. The SNF can be diverted toa nuclear weapon for military purpose, hence, the accurate interrogation
technique has been required to prevent missing, theft, or defect of the SNF. For
pin-by-pin monitoring of SNF, gamma emission tomography (GET) device,
which named Yonsei single photon emission computed tomography version 3(YSECT.v.3), was proposed and optimally designed in our previous study. In thispaper, we evaluated the performance of YSECT.v.3 device, and developed amachine learning-based denoising technique in order to improve the inspection
accuracy of detecting partial defects.
The YSECT.v.3 device was composed of tungsten collimator, 64-channel GAGG
scintillator, and silicon multiplier. For validation of the performance of fabricateddevice, the projection images and sinograms were obtained using the 1 μCi Cs137 sources. The sinograms were converted to the tomographic images with theiterative reconstruction algorithm. However, the central region of tomographic
image showed low-image quality compared with the external region because ofhigh-attenuation and -scatter condition for interrogating the partial defects
within the SNF. For this reason, the machine learning-based denoising technique
was developed. The 5,950 tomographic images, which acquired using the
GEANT4 toolkit, were used for training, and 50 images were employed to thedenoising technique for the test. The performance of the YSECT.v.3 device was207Abstracts - Oral

O7 - 04

evaluated for spatial resolution of projection images and tomographic images,
and the performance of the denoising technique was validated through patternclassification accuracy.
The minimum pitch size between each fuel rod in the PWR type SNF used in
Korea is 12.32 mm. The spatial resolution of the YSECT.v.3 device was defined
through FWHM on the projection and tomographic images. The convolutional
autoencoder (CAE), which is an artificial neural network optimized to image data,
was employed for denoising technique. The developed CAE algorithm was asymmetric structure, and it was composed of the input layer, encoder (2xconvolution layer, 2x maxpooling layer), decoder (2x upsampling layer, 2x
maxpooling layer), and the output layer. The pattern classification accuracy ofdenoising image was analyzed as almost 100%.
In this study, we developed machine learning-based denoising algorithm andevaluate the feasibility to the prototype GET device (YSECT.v.3). The current
results indicate denoising algorithm could improve the inspection accuracy. Our
further study will include the experimental evaluation for YSECT.v.3 device withvarious measurement conditions.
Acknowledgement
This research was supported by the Nuclear Safety Research Program throughthe Korea Foundation Of Nuclear Safety (KoFONS) using financial resources
granted by the Nuclear Safety and Security Commission (NSSC) of the Republicof Korea (No. 2106073); the Basic Science Research Program through the
National Research Foundation of Korea (NRF) funded by the Ministry of Education(NRF-2021R1I1A1A0105987512); the Korea Institute of Energy Technology
Evaluation and Planning (KETEP), and the Ministry of Trade, Industry & Energy(MOTIE) of the Republic of Korea (No. G032579811).
Reference
E. A. Miller et al., “Hybrid Gama Emission Tomography (HGET): FY16 Annual
Report,” 2017. L. E. Smith and P. Jansson, “A Viability Study of Gamma Emission
Tomography for Spent Fuel Verification : JNT 1955 Phase I Technical Report,” no.
October, p. 119, 2016. H. joo Choi et al., “Experimental evaluation of fuel rod
pattern analysis in fuel assembly using Yonsei single-photon emission computed
tomography (YSECT),” Nucl. Eng. Technol., vol. 54, no. 6, pp. 1982–1990, 2022.

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O8 - 01#8 Radioactive Waste and Current Radiological Issues

Factor Analysis of In-Situ Object Counting System(ISOCS) for
Uranium Quantification in Radioactive Waste Drums

Jihyun Ahn1, Seungmin Lee2, Hee Seo1,*
School of Jeonbuk National University/Korea1, Department of Korea Institute of
Nuclear Nonproliferation and Control/Korea2
Email: hseo@jbnu.ac.kr
Keywords: Safeguards, Gamma Spectroscopy, ISOCS, Radioactive Waste,
MCNP
The international community should adhere to safeguards agreements to ensure
the peaceful use of nuclear energy. Quantifying uranium in radioactive waste
drums is an essential part of safeguards activities. Destructive assay (DA) andnon-destructive assay (NDA) techniques are commonly used; however, the DA
is costly and time-consuming. On the other hand, In Situ Object Counting System(ISOCS) is an NDA technique that allows for efficiency calibration without
standard materials, making it efficient and cost-effective.
In this study, we evaluate the performance of ISOCS for quantifying uranium inconcrete radioactive waste drums. To this end, an experimental drum at KEPCO
NF and low-level radioactive waste (LLRW) drums at the Korea Radioactive Waste
Agency (KORAD) were used. The study compares the results obtained fromISOCS to those obtained through Multi-Group Analysis for Uranium (MGAU) and
DA. The factors that affect uranium quantification, including detection geometry,
spectrum quality, heterogeneity, and self-shielding, were examined by using
MCNP simulations.
The results showed that the ISOCS could quantify uranium in radioactive waste
drums, with a relative error ranging from -1% to -67% compared to DA results.
These results suggest that, depending on the measurement conditions, theISOCS results are comparable to the DA results. Also, the ISOCS results were
found to be more accurate than those obtained from the MGAU application. By
using the MCNP simulations, it was found that the geometrical condition of thedrum had a negligible influence on the ISOCS results and spectrum quality was
found to depend on measurement time. Heterogeneity was a significant factor inanalyzing LLRW drums, and the self-shielding effect increased as the total
amount of uranium increased.
Overall, these results indicate that the ISOCS method can be used for safeguards
activities as it can provide accurate results while minimizing the complex

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processes required for uranium quantification in radioactive waste drums. The
study also highlights the importance of selecting appropriate methods foruranium quantification based on measurement conditions and sample
characteristics. The study's findings provide valuable insights into the factors
affecting uranium quantification and can be used to develop an improved nuclear
material measurement technology for safeguards purposes.
Acknowledgement
This study was supported by the National Research Foundation of Korea (NRF)
funded by the Korea Customs Service and Ministry of Science and ICT (NRF2021M3I1A1097913). This study also was supported by the Nuclear Safety
Research Program through the Korea Foundation Of Nuclear Safety (KoFONS)
using financial resources granted by the Nuclear Safety and Security Commission(NSSC) of the Republic of Korea (2004024-0322-CG100); further support wasreceived from the Korea Evaluation Institute of Industrial Technology (KEIT)
funded by the Ministry of the Interior and Safety (No. 20014778). This research
was part of the project titled ‘Development of Smart Processing Technology for
Sea Foods,’ funded by the Ministry of Oceans and Fisheries, Republic of Korea.

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O8 - 02#8 Radioactive Waste and Current Radiological Issues#8 Radioactive Waste and Current Radiological Issues

Development of AI-based Anomaly Detection Model for Nuclear
Waste Data

Jinhyung Kwon, Young-su Kim, Jiseok Kim, Hanjin Kim, Donghyeok Shin,Yongkyun Kim
Korea Atomic Energy Research Institute (KAERI), Daejeon, Korea1, Departmentof Nuclear Engineering, Hanyang University, Seoul, Korea2
Email:
Keywords: Aritifitial intelligence
Nuclear waste is a product of nuclear power operation that emits radiation and it
must be stored, managed, and disposed properly for safety. For conducting theproper process of radioactive waste management, it is essential to classify it
through the concentration of nuclides to prevent damage to humans (Corkhill andHyatt et al., 2018). However, there has been increasing concerns regarding data
errors, especially human error, that occurred during the analysis of low- andintermediate-level nuclear waste (Chung et al., 2020).
In recent years, the use of artificial intelligence (AI) technique has received
significant attention applied across many domains to solve complex problems
including nuclear power industries. The sheer volume of data collected in nuclear
waste facilities makes them valuable locations for incorporating AI-based
technologies.
In this paper, we provide an AI based anomaly detection algorithm for sustainable
management of low-intermediate nuclear waste that is expected to increase in
the future. Activity concentration data of radionuclides in nuclear waste were
produced using scaling factor of Korea hydro & Nuclear Power (KHNP) andElectric Power Research Institute (EPRI), and deep neural network was trained byproduced dataset. As results, the AI-based anomaly detection system to ensure
the accuracy of nuclide concentration data was developed. The performance of
the anomaly detection algorithm was evaluated for various human error casesthat can happen in the process of nuclear waste characterization, and theapplicable concepts were investigated.
The developed model is expected to perform for rapid decision making on humanerror of radioactive waste data that are generated during operation anddecommissioning of nuclear power facilities.

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Abstracts - Oral

O8 - 03#8 Radioactive Waste and Current Radiological Issues

Research on the key problems of the new smelting andpurification process of metal radioactive pollutants

Kewei Jiang
China Institute for Radiation Protection1
Email:
Keywords: Nuclear decommissioning, decontamination, smelting, volume reduction, radioactive scrap
By setting up a new intermediate frequency induction furnace melting system,
matching the negative pressure dust removal chamber and radioactive exhaust
treatment system, this project determines the improved and optimized deconfouling production and scientific research technology compared with the originaltechnical scheme, and improves a complete set of intermediate frequency induction furnace melting deconfouling equipment, which is suitable for the rescientific research and reuse of radioactive ingot. In this project, molten salt is
added as the medium. By adopting electrochemical technology, high-temperature molten steel is used as the liquid cathode, and high-temperature resistantmolybdenum sheet is used as the anode. The potential difference is used to
selectively separate residual nuclide Limited company with similar properties. Improving the application of technology and technology can reduce the abandonment rate of different metals, improve the casting production quality control levelof nuclear industry, ensure the stability and standard of product quality, and havecertain reference significance for the development of metal radioactive waste
smelting and pollution removal technology route and ideas to improve.

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#8 Radioactive Waste and Current Radiological Issues#8 Radioactive Waste and Current Radiological Issues

The Study of Mechanical wear principle and dynamics of graphite
spheres in HTR-PM high temperature gas cooled reactor

Lixiao Guo
China Institute for Radiation Protection1
Email:
Keywords: Mechanical wear principle,dynamics study,graphite spheres
The load and temperature are the key factors affecting the friction and wear rateof graphite balls in the core of sphere-bed HTGR. A self-designed friction andwear testing machine was used to study the friction loss factors and friction
coefficient of graphite ball used in HTR-PM high temperature gas-cooled reactor
under different atmosphere under load. The relationship between the wear rateof graphite ball and the load and temperature of the graphite ball was obtained
through fitting, so as to determine the wear mechanism. By fitting theexperimental data, the relationship between wear rate and load is exponential,
and logarithmic with temperature. Two influencing factors were investigated at
the same time, and the formula of wear rate was as follows :q = 16.55ln T +5.99e0.0036F-48.57. Combined with the load and temperature distribution ofgraphite spheres in the core of HTR-PM, the amount of stone and ink dust
produced by the friction between graphite spheres in the core is about 14.01
g/d(5.1 kg/a). Combined with the load and temperature distribution of fuelelements in HTR-PM demonstration reactor, it is calculated that the amount ofgraphite dust produced by friction between graphite pellets is about 14.01
g/d(5.1 kg/a).

213

Abstracts - Oral

O8 - 05#8 Radioactive Waste and Current Radiological Issues

Development of Microwave Drying Technology for Radioactive
Wet Waste Treatment in Drum

Meng Zihe1,*, Gao Chao1, Jia Meilan1, Liang Dong1
The Department of Radioactive Waste Management, China Institue for Radiation
Protection, China1
Email: 1499095216@qq.com
Keywords: Microwave Dying, Wet Waste Treatment, Radioactive Waste
During the development of nuclear energy and the application of nuclear
technology, a large amount of low-to-intermediate level radioactive waste is
generated. For wet waste containing a high amount of water such as concentrate,
waste resins, sludge, and cleaning cloths, there is a problem with an increase inwaste volume if they are directly treated by cement solidification. Therefore, awet waste management plan involving dewatering/drying, preparation, temporarystorage, and disposal has attracted increasing attention. Drying is a key part ofthis management plan, and through drying, the form of various types of wet waste
can be transformed. Microwave heating technology has the characteristics ofhigh energy utilization, low thermal inertia, fast heating rate and uniform heating,and has been widely used in the field of drying.
Microwave drum drying technology can achieve volume reduction of wet wastematerials and concentrate most of the radioactive nuclides in the drum. Basedon the study of the physical properties and the microwave drying characteristics
of various wet waste, microwave drum drying devices can develop different
drying processes to achieve drying of various wet waste. This microwave drum
drying device includes a feeding unit, a microwave heating unit, a cooling system,
an exhaust gas treatment system, and auxiliary drying devices.
According to the experiment results, the microwave drum drying device with amicrowave power of 12 kW can achieve an average evaporation rate of more than
6.0-7.0 kg/h, and a product moisture content of 1-2%. For the concentrate andsludge with a water content greater than 60%, the capacity reduction factor isgreater than or equal to 2.0. For the cleaning cloths with a moisture content of
60%, through enhanced ventilation measures, it can be dried below 100 ℃ of thecleaning cloth, and the moisture content of the product is 10%.

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#8 Radioactive Waste and Current Radiological Issues#8 Radioactive Waste and Current Radiological Issues

Development and verification of portable incineration technology
for low-level radioactive waste

Chu Haoran1,*, Xu Wei1, Zheng Bowen1, Zhang yu1, Ruan Jiasheng1
China Institute for Radiation Protection，Taiyuan，China1
Email: chuhaoran@cirp.org.cn
Keywords: pyrolysis incineration, low-level radioactive waste, portable
Several fixed incineration facilities have been built in China, achieving volume
reduction and stabilization treatment for low level radioactive combustible solid
waste. However, for research institutes and small nuclear facilities those are faraway and produce small amount of waste, it is not economical to build fixedincineration facilities. Portable incineration devices can be transported by trucks
to the waste generation or temporary storage sites, and then incinerate to reducethe volume of waste. Therefore, the waste transportation process is no longer
required, and the equipment manufacturing costs are also reduced.
Based on the development and application of fixed incineration technology,
China Institute for Radiation Protection developed and verified the portable incineration technology for low-level radioactive waste. Firstly, a portable incineration test device in a 40-foot ISO standard container was established by optimizing the pyrolysis and incineration process and miniaturizing equipment. Secondly, the device was tested and verified with simulated waste and real radioactive waste, respectively. Finally, the safety requirements and evaluation principles
were studied to ensure that the transportation of the device after use is feasible.
The verification results show that the average treatment capacity for combustible
solid waste is 26.19kg/h, the average volume reduction ratio of radioactive waste
after incineration is 47.0, the emission limit of main pollutants in off gas meetsthe " Standard for pollution control on hazardous waste incineration ", the clean
coefficient of nuclides is 1.3E6, and the average loss on ignition of ash is 3.63%.
The portable incineration device decontaminated after use can be transported bytruck as surface contaminated objects. It is believed that under normal transportation conditions, it will not cause unacceptable impacts on staff, the public, andthe environment, and the radiation risk is controllable.
Compared to fixed incineration facility, portable incineration device can achieve
the same waste treatment effect with smaller floor space and fewer operating
personnel, greatly reducing construction and operating costs. Portable incineration technology and fixed incineration technology complement each other, which
can meet more waste treatment needs and provide a more economical and feasible way for low level combustible waste treatment.
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Abstracts - Oral

O8 - 07#8 Radioactive Waste and Current Radiological Issues

Research on Si-Al base Geopolymers Formulation of SimulatedRadioactive Fluoride Liquid Wastes

Li Zhang, Xiliang Guo
China Institute for Radiation Protection1
Email:
Keywords: Fluoride liquid waste, Geopolymer solidification, Fluoride ion
leaching
The radioactive fluoride liquid wastes produced during the operation of thoriumbased molten salt reactors is both radioactive and chemically toxic, and needs
to be solidified to meet the requirements for safe disposal.
In this paper, Na2O·nSiO2 and fly ash were used as the solidification substrates,
and the simulated radioactive fluoride liquid wastes was solidified with Si-Al basegeopolymers by carrying out orthogonal experiments of four factors and fourlevels.
The results show that when Na2O·nSiO2 module is 1.0, the solid-liquid ratio is0.40, and the alkali initiator content is 20%, the simulated radioactive fluoride
liquid wastes can be effectively solidified, and the 28-day compressive strengths,
immersion resistance and simulated nuclide leaching rate of the solidified bodymeet the relevant requirements of “Performancerequirements for low andintermediate level radioactive waste form－Cemented waste form”(GB 14569.12011), and leaching rate of fluoride ion meets the relevant requirements of“Identification standards for hazardous wastes-Identification for extraction
toxicity”(GB 5086.3-2007).

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#8 Radioactive Waste and Current Radiological Issues#8 Radioactive Waste and Current Radiological Issues

The Study on Microwave Drum Drying for Radioactive Wet Waste

Gao Chao1, Jia Meilan1, Meng Zihe1, Yan Xiaojun1,*
Radioactive Waste Management Department, China Institute for
Radiation Protection, China1
Email: top-super-007@163.com
Keywords: microwave;drum drying;radioactive waste;wet waste;concentrated
liquid;slurry
Purpose: At present, the radioactive wet wastes from NPPs in China are mostly
treated by temporary storage and cement solidification, but the capacity of cement solidification is larger, generally up to 4～6, which not only occupies precious temporary storage space, increases the burden of final disposal in future,
but also has some potential safety hazards, and does not conform to the principleof waste minimization. Therefore, IAEA-TECDOC-1504 "Innovative Technology
for Radioactive Waste Disposal and Completion in Nuclear Power Stations" andothers recommend new treatment technologies for radioactive waste. Many
countries in the world have carried out research on the treatment of radioactive
waste by microwave. Taking the most extensive, in-drum microwave drying technology as an example, since 1950s, Oak Ridge National Laboratory, Departmentof Energy (United States, DOE), Kraftanlogen Heidelberg Company, Lin High
Therm Company and Nukem Company of Germany have successively carried out
the research on microwave drum drying process and equipment and obtained theresults. The microwave drum drying technology has been applied to treat NaClslurry, TRU slurry and spent resin in stainless steel containers. Some research
results have reached pilot scale. Although different countries have different research and treatment objects, microwave is far superior to other common treatment methods in terms of capacity and effect. Research on Microwave Technology for radioactive waste treatment has a positive impact on waste treatment ofdomestic nuclear facilities. Microwave treatment is a new type of radioactive
waste treatment technology which can minimize the volume of waste.
Methods:In this paper, the microwave drum drying process of two kinds of radioactive wastes, i.e. concentrated liquid and slurry, have been studied. Feasibility
study, bench test study and expanded test study have been carried out respectively. The control parameters of 12L bench and 200L bench(Project scale) for
concentrated liquid and slurry have been determined. The water content, volume
reduction ratio and water evaporation rate of the products were determined.

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Results: For liquid waste, 713.71kg concentrated liquid was added in this drying
test, the total drying time was 216300s, and the average evaporation rate was
6.70-6.93kg/h; A barrel of dry products with a filling rate of 85% is obtained, andthe volume reduction coefficient is 3.5; The dried product is 245kg, without freewater, with an average moisture content of 1.63 ~ 3.32%. The dried product of
the simulated concentrate is uniform as a whole, with uniform moisture content
distribution and good consistency. For slurry, The dried product is dark reddish
brown dry solid, without free water, and there are cracks on the surface andinside. After the barrel is cut open, it is a loose block solid; The measured moisture content of the dried product is 0.51% - 8.91%; The evaporation rate per unit
time is 5.90kg/h, which is lower than the evaporation rate per unit time of thesimulated concentrate, which may be related to the difference in the properties
and moisture content of the two materials; For simulated slurries with different
amounts of materials, under the condition of sufficient microwave power, they
can be dried to the level of low water content, which can reach the lowest levelof nearly 0. However, from the perspective of actual work efficiency and subsequent disposal requirements, there is still much room to relax the control level ofthe moisture content of the dried products of the mud, so that the evaporation
rate per unit time can be further improved.
Conclusion:Microwave drum drying process of concentrated liquid and slurry is
feasible. Its drying products are uniform, the process is controllable, and thetreatment efficiency is high. It is an effective means to solve the volume reduction
treatment of wet waste.

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#8 Radioactive Waste and Current Radiological Issues#8 Radioactive Waste and Current Radiological Issues

A synthesized bifunctional strongly basic anion exchange resinfor ReO4-/TcO4- removal

Yujia Kang1, Ping Xu1, Huiping Tang1, Shiquan Cao1, Huiyun Xue1, Zhi Chen1,*
School of Nuclear Science and Technology, University of Science andTechnology of China, Hefei, China1
Email: zchen@ustc.edu.cn
Keywords: pertechnenate, perrhenate, removal, anion-exchange resin
Technetium-99 (99Tc) is the most significant long-lived fission productproduced by nuclear fission of 235U with a notable fission yield of 6.0507 %.
Owing to its long half-life of 211000 a, facility to form an anionic state as TcO4-,
and high mobility of pertechnetate in groundwater, the management of 99Tc has
become a critical consideration in the disposal of high-level radioactive waste
liquid. Hence, it is essential to remove 99Tc from the radioactive wastewater.
Resin-based ion exchange methods have been widely used to remove 99Tc.
In this study, a bifunctional strongly basic anion exchange resin with asymmetricstructure was synthesized via the reaction of chloromethylated styrene–
divinylbenzene copolymer with N, N-dimethylhexylamine to carry out the removal
work. Different analyses including scanning electron microscopy (SEM), Fouriertransform infrared spectroscopy (FITR), and X-ray photoelectron spectroscopy(XPS) were adopted for characterizing the microscopic morphology and structureof the synthesized resin. Considering the radioactivity of all isotopes of Tc, thenon-radioactive ReO4- as an analogue for TcO4- was used to conduct staticadsorption experiments. The adsorption behaviours of the synthesized resin forthe separation of ReO4- were investigated as functions of temperature, solution
pH, initial concentration, the mass of resin, and contact time. To further
understand the adsorption mechanism, we also took into account both sorption
thermodynamics and sorption kinetics.
Microscopic characterization results showed that the surface of the synthesizedresin became rough compared to the resin before adsorption, indicating that
ReO4- could be adsorbed onto the resin. The batch sorption tests demonstrated
that the synthesized resin showed an excellent adsorption capacity for ReO4over a wide pH range of 3~11 as well as rapid adsorption kinetics. The adsorption
isotherm was found to fit the Langmuir model more effectively (R2>0.9), whichindicated that the perrhenate sorption on the synthesized resin was monolayer
adsorption. Furthermore, the sorption kinetics results showed that the adsorption219

Abstracts - Oral

O8 - 09

process followed the pseudo-second-order kinetics model. This suggested thatthe rate-controlling process was governed by the chemical reaction for ion
exchange. Compared to the currently available single-functional group anion
exchange resins used for ReO4- removal, the synthesized resin exhibited higher
adsorption capacity and faster removal efficiency, which was attributed to theasymmetric functional groups with longer alkyl chains that enhance theadsorption selectivity while the shorter alkyl chains maintain the adsorption
capacity and exchange kinetics. Besides, the synthesized modified resin
demonstrated excellent cyclic performance.
Overall, the synthesized resin has demonstrated exceptional removal efficiency
for ReO4-, which endows the resin with the ability to adsorb TcO4-, making it apromising candidate for practical use in the removal of radioactive 99Tc in theenvironment.

220

O8 - 10#8 Radioactive Waste and Current Radiological Issues#8 Radioactive Waste and Current Radiological Issues

Research and application of on-site calibration technology for aradioactive waste drum NDA inspection system

Hengguan YI1,*, Zhigang LI1, Zhihui TANG1, Zhenglin HUANG1, Yingjing WEI1,
Qinjian CAO1
Shanxi Provincial Key Laboratory for Radiation Safety and Protection, Taiyuan,
China & Key Laboratory of Radiation Protection of CNNC, Taiyuan, China &China Institute for Radiation Protection, Taiyuan, China1
Email: kidyhg1412@sina.com
Keywords: radioactive waste drum inspection system, non-destructive assay,
on-site calibration, HPGe detector characterization, Geant4 package
The measurement and management of radioactive waste is strongly related tothe safety and economic issues of nuclear power plants. It is significant forradioactive contamination prevention and control, protecting people and theenvironment, and nuclear energy and technology promotion and utilization.
During the operation of nuclear power plants, different types of contaminatedradioactive wastes have been generated. According to the latest regulations suchas IAEA safety standards series No. GSG-1 classification of radioactive waste,
the majority of them are very low level waste(VLLW), low level waste(LLW) andintermediate level waste (ILW), which are generally stored in radioactive waste
drums in a radioactive waste repository for interim storage after certain
processing. Under China’s laws and regulations, the solid radioactive waste will
be sent to a disposal facility for final disposal. Before storage or disposal, thekinds of naturally occurring radionuclides and artificial radionuclides with levelsof activity concentrations must be identified and analyzed. Currently, radioactive
waste drum inspection system are widely used to scan the solid radioactive waste
outside a 400 L drum with non-destructive assay (NDA) methods in China, suchas Fujian Ningde Nuclear Power Company Limited, Liao Ning Hong Yan He
Nuclear Power Company Limited and Yangjiang Nuclear Power Company Limited,
etc. A HPGe detector is utilized to collect the gamma spectrum of radioactive
waste, and several optional doserate meters are also used to measure the dose
rate at the surface of the container and at 1 meter therefrom. Because of thecomplicate routine maintenance and the highly professional data analysis, there
are various problems in running and using these equipment. It is difficult forworkers to make full use of these equipment to properly evaluate the levels of221

Abstracts - Oral

O8 - 10

activity concentrations of solid radioactive waste in a drum and then determine
the class of the waste.
For this reason, we have developed an on-site calibrate method and designed aunique radioactive source holder to test the HPGe detector in a lead collimatorand several doserate meters on-site. An accurate Monte Carlo numerical modelof the HPGe detector has been established using Geant4 package, and thelocations and thicknesses of the HPGe crystal dead layer have been carried out.
The numerical model of radioactive waste drum inspection system with a varietyof drums therefore can be built. With an approximate model of the measured
waste drum under the assumption of homogeneity, the activity concentration,
uncertainty and maximum relative deviation of selected radionuclides have been
calculated and estimated.

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O8 - 11#8 Radioactive Waste and Current Radiological Issues#8 Radioactive Waste and Current Radiological Issues

Activation status of metallic components in the PET-cyclotron
revealed by destructive and non-destructive analysis

Go YOSHIDA1,*, Hiroshi MATSUMURA1, Akihiro TOYODA1, Hajime NAKAMURA1,
Kazuyoshi MASUMOTO1, Taichi MIURA1, Iwane KOUMURA2, Hiroyuki UNO2,
Masahiko KOZAKI3, Ayumi YAMAMOTO3High Energy Accelerator Research Organization (KEK), Tsukuba-Ibaraki,
Japan1, Sumitomo Heavy Industries, Ltd., Tokyo, Japan2, SHI-ATEX Co., Ltd.,
Seika-Kyoto, Japan3, Tokyo Metropolitan Geriatric Medical Center, Tokyo,
Japan4
Email: yoshigo@post.kek.jp
Keywords: Activation Accelerator Decommissioning
Along with the positron emission tomography (PET) began to become popular inJapan since 1990s, the number of cyclotrons for PET pharmaceutical production(PET-cyclotron) has increased rapidly. Cyclotrons installed in the early years
have been subject to decommissioning, and the number has been gradually
increasing. Normally accelerators become activated with operation and thedegree of activation determines the amount of radioactive waste at thedecommissioning, thus evaluation of them is crucial. However, an effective
evaluation method has not been established yet. Currently, a main-body ofcyclotron is disposed as radioactive waste and it cost lots of money. In this study,
we clarified the activation status of iron yoke, which is one of the heaviest
components of the cyclotron, by destructive analysis with core boring. Nondestructive analysis using a γ-ray detector were also conducted in parallel, andoptimization of actual decommissioning were discussed.
Tokyo Metropolitan Geriatric Hospital, which the PET-cyclotron had operated
until 2013, was selected as a target facility. More than 6 years had passed until
our survey began in 2019. First, we investigated surface distribution of induced
activities using a CsI survey meter and a CeBr3 spectrometer combined with theshields. Shielding is useful in an environment such as a PET-cyclotron room
where the entire room is activated, it can be reduced the influence of radiationfrom outside the object.
We found 60Co on the almost area of the yoke and 54Mn near the target as
residual nuclides. We also investigated the center-pole where protons areaccelerated and revealed dose distribution in detail. We also determined thedepth distribution of activity by core samples taken from four representative

223

Abstracts - Oral

O8 - 11points on the side and top of the yoke. It was found that the 60Co activity
decreased exponentially from the inside to the outside of the accelerator andturned to increase again in the region about to the outside surface. The resultsof CeBr3 reproduced that of the core sample in the equivalent region, and thelinearity with the surface dose-rate was also confirmed. Possibility for the nondestructive estimation method for the 60Co activity inside the yoke wassuggested.
Acknowledgement
Authors addition Katsuhiko SAITO, (Aff. 2), katsuhiko.saito@shi-g.com Kei
WAGATSUMA, (Aff. 4),

224

#8 Radioactive Waste and Current Radiological Issues

O8 - 12#8 Radioactive Waste and Current Radiological Issues

Efficient Adsorption of Sr2+ from High Level Liquid Waste by 4A
Zeolitexiang qin1, chao gao1,*, xiliang guo1, xia tao1
China Institute for Radiation Protection/china1
Email: cau-227@163.com
Keywords: 4A zeolite, high-level liquid waste, ion exchange, Sr2+
The efficient separation of 90Sr from high level liquid waste generated from the
PUREX process for treating spent fuel is of great significance for reducing thedisposal cost of high level liquid waste and improving the disposal safety. In thiswork, we treated the Sr2+concentration by precipitation separation and ion
exchange (4A zeolite) to 48.5 ppm (activity concentration equivalent to
2.5×1011Bq/L), free nitric acid concentration of 2mol/L, high concentration of
NaNO3, Fe (NO3)3, AlNO3, Cr (NO3)3, Ni (NO3)2, KNO3, and CsNO3 (1.24mol/L
Na+, 0.22mol/L Fe3+, 0.1mol/L Al3+, 0.02mol/L Cr3+, 0.05mol/L Ni2+,
0.02mol/L K+, Na+/Sr2+=587.63,; K+/Sr2+=14.43). It was found that 4A zeolite
had a significant effect on Sr2+in the simulated high-level liquid waste. It has
high separation efficiency and selectivity, with an adsorption rate of 93%, The
distribution coefficient Kd is as high as 1298.47 mL/g. At the same time, 4A
zeolite exhibits fast adsorption rate and efficient adsorption kinetics (about 2h),high capture capacity (294.12 mg/g) in high salt and complex environments, awide pH range (4-9), and excellent radiation resistance. After irradiation with
1000KGy-60Co, the crystallinity and properties of 4A zeolite were not affected.
The ion exchange experiment shows that the replacement order of 4A zeolite fordifferent ions in high level liquid waste is Cr3+>Fe3+>Ni2+>Sr2+>K+>Na+>Cs+.
These superior characteristics, coupled with the characteristics of simple
synthesis methods, low cost, and environmental friendliness, make 4A zeolite
highly competitive in the field of Sr separation in high-level liquid waste.

225

Abstracts - Oral

O8 - 13#8 Radioactive Waste and Current Radiological Issues

Dose estimation for radioactive metal waste handling workers
using dose conversion factors of ICRP Publications 119, 141, and
EPA FGR No. 11

Jin Ho SON1, Tae Young KONG1,*
Department of Nuclear Engineering, Chosun University, Gwangju/Korea1
Email: tykong@chosun.ac.kr
Keywords: Dose Estimation, Dose Conversion Factors, Radioactive Waste
System decontamination or cutting metals and concrete in nuclear power plants
causes occupational exposure during decommissioning. To reduce occupational
exposure, various radiation protection measures should be prepared byestimating the exposure dose in advance. RESRAD-RECYCLE, the computercode developed by the Argonne National Laboratory in the US, is generally used
for estimating occupational dose when handling metals contaminated withradioactive materials. RESRAD-RECYCLE used the dose conversion factors(DCFs) of EPA FGR No. 11 based on ICRP Publications 30 and 48 published in
the 1980s for internal exposure estimation.
This study compared the DCFs of EPA FGR No.11 used by RESRAD-RECYCLE
with ICRP Publications 119 and 141. In addition, the exposure doses of radiation
workers were calculated and compared by changing the DCFs in the library of
RESRAD-RECYCLE to the DCFs presented in ICRP Publications 119 and 141.
ICRP Publication 119 provides DCFs of inhalation and ingestion based on the1990 Recommendations of the ICRP. ICRP Publication 141 also provides DCFs
based on the 2007 Recommendations of the ICRP. The results of the DCF
comparison between EPA FGR No.11 and two ICRP Publications showed that
DCFs are similar for gamma-ray emission nuclides, but DCFs for beta-ray
emission nuclides in ICRP Publications are much lower than those in EPA FGR
No.11.
This study also estimated the exposure dose of radioactive metal waste handlingworkers in Hanul Units 1 and 3 and Hanbit Unit 4 under the assumption that thosereactors would be decommissioning in the future. The expected radioactivity
concentration of the steam generator chamber of each unit was used as thesource term for occupational dose estimation. As a result, it was found that the
exposure doses of radiation workers who handled steam generator wastes using226

O8 - 13#8 Radioactive Waste and Current Radiological Issues

DCFs of ICRP Publications 119 and 141 were a little lower than the exposure
dose using DCFs of EPA FGR No.11.
Acknowledgement
This research was supported by the Nuclear Safety Research Program throughthe Korea Foundation of Nuclear Safety using the financial resource granted by
the Nuclear Safety and Security Commission of the Republic of Korea (No.
2106024-0323-SB110).

227

Abstracts - Oral

O8 - 14#8 Radioactive Waste and Current Radiological Issues

Speciation simulation and sorption mechanism of 238Pu in waste
repository groundwater

CHEN Chao, ZHANG Aiming, LUO Zhiping*
Department of Nuclear Environmental Science, China Institute for Radiation
Protection, China1, Department of Radiation Safety, China Institute of Atomic
Energy, China2
Email: luozhiping00@163.com
Keywords: radionuclide; irreversible adsorption; migration; species distribution
The assessing of the mobility and fate of radionuclides escaping from radioactive
waste disposal facilities in the environment relies on the use of numerical models
capable of simulating all of the important processes along their complex
geological environment. Adsorption models usually subsumes processes such as
element chemistry and their chemical reactions with the surrounding environment,
ion exchange or incorporation processes which have to be considered in longterm safety assessments for radioactive waste repositories. pH and redox
conditions are recognized as the most important parameters that influence these
sorption and migration processes. The studies on the batch experiments werecarried out under different host rock formations for waste repositories, atheoretical model of nuclide migration in soil and groundwater environment wasconstructed, and the irreversible adsorption reaction mechanism of the microinterface was studied through static adsorption-desorption data. And speciation
distribution of 238Pu in groundwater systems were calculated, using PHREEQC
software with the recent thermodynamic data published by OECD/NEA tounderstanding the sorption, diffusion and migration behaviors.
The static adsorption and desorption experiments adopt batch method, and thetest medium different host rock formations for waste repository from an arid
mining area in Gansu Province, China. According to the field measurement, thepH value in the groundwater was 7.48 and Eh was 125.7mV. The solid-liquid ratioof 1:10 g/mL was added into the centrifuge tube, and the nuclide activity
concentration was measured after adsorption and desorption equilibrium toobtain the adsorption and desorption isotherms. The PHREEQC software wasfurther used to study the speciation changes of 238Pu in the groundwater of the
pre-selected area of the waste repository under the conditions of changing pH
and redox conditions. The results show that the absorption isotherms of 238Pu
doesn’t coincide with its desorption isotherms. According to the metastable228

O8 - 14#8 Radioactive Waste and Current Radiological Issues

equilibrium adsorption theory, 238Pu has an irreversible adsorption-desorption
process, ion exchange and surface complexing are the main way of adsorption.
The adsorption isothermal line is basically in accordance with the Freundlich
adsorption isotherm. Utilizing the Equilibrium program in PHREEQC software
package, the species and concentration-controlling factor of 238Pu in waste
disposal facilities groundwater were investigated .Results indicate that Pu(OH)4,
Pu(SO4)2-, and PuSO4+ are the predominant species in this groundwater .The
results also indicate that 238Pu concentration in this groundwater is mainly
controlled by the precipitation of saturated minerals as Pu(OH)4. When keeping
Eh constant at 125.7mV, 238Pu concentration is controlled by the mainly speciesof elemental Pu(SO4)2-, PuSO4+, Pu3+, Pu(OH)4, Pu(OH)3+ and Pu(OH)22+
with the changing of pH condition that the groundwater composition is notchanged .When keeping pH constant at 7.48, the mainly species will bePu(SO4)2-, PuSO4+, Pu(OH)4, PuO2OH and PuO2OH+ with the changing of Ehcondition.

229

Abstracts - Oral

O8 - 15#8 Radioactive Waste and Current Radiological Issues

Gaseous radioactive iodine capture of microporous molecularsieves supported silver: BEA, MFI and CHA

Zexiang Chen1, Jie Yu1, Xin Li1, Jianlu Pei1, Shuwei Liang1, Yongguo Li1,*
Key Laboratory of Nuclear Air Purification Technology, China Institute for
Radiation Protection, China1
Email: lyg0624@163.com
Keywords: radioactive iodine, silver zeolite, microporous molecular sieves, CH3I
captureThe efficient capture of radioactive fission products in the course of a severe
nuclear accident is a major issue for safe nuclear energy. Gaseous radioactive
iodine compounds are particularly paid attention due to their large abundance,long half-life and human pathogenicity. Silver zeolite was one of the mostefficient sorbents for the capture of iodine compounds in both severe nuclear
accidents and reprocessing of nuclear fuel.
Previously, most of past works focused on silver faujasite zeolites (namely X type)and especially silver mordenite zeolites with little attention paid to other types ofzeolites. Nevertheless, several important issues remained to be solved about theeffect of zeolitic parameters on their adsorption behavior, with a special
emphasis to be placed on CH3I. This paper is concerned with the capture
performance of gaseous CH3I in microporous molecular sieves supported silver(BEA, MFI and CHA), which are typically used in petroleum industries with certain
pore sizes (5.95 Å, 4.7 Å and 3.72 Å) close to the kinetic diameter of CH3I
molecule (4.23Å).
11.8 wt% silver was loaded on the three types of zeolites by impregnation method.
XRD and BET results showed that the crystal phase and pore size distribution of
three types of zeolites kept intact after Ag loading, although the relative
crystallinity and specific surface area dropped 30-40% and 15-25%, respectively.
Quantification by H2-TPR combined with the XPS observation showed that all
silver existed in the form of Ag+ ions. However, the redox properties of Ag overthree types of zeolites differed. Ag+ on Beta zeolite displayed the most active
redox ability. The affinity of Ag to CH3I was calculated by the DFT method. The
bond energy between Ag and I was highly correlated with the redox ability of Ag+
ions. The bond energies of Ag-I for I2 and CH3I over Ag/Beta were as high as
222.8 and 189.4 kJ/mol, respectively. The storage capacities of CH3I for Ag/Beta,
Ag/ZSM-5 and Ag/SSZ-13 were measured as 208 μmol/g, 90 μmol/g and 158

230

O8 - 15#8 Radioactive Waste and Current Radiological Issues

μmol/g, respectively. It can be found that the pore size with a span of 3.72-5.95
Å had little effect on the capture performance of CH3I. Importantly, the redox
ability of Ag+ on different zeolites determined its affinity with CH3I, which
ultimately affected the storage capacity of CH3I.
This work expands the scope to the field of silver-based microporous molecularsieves. The good thermal stability of microporous molecular sieves will endow
Ag-based iodine adsorption materials with more application value under severe
nuclear accidents.
Acknowledgement
We acknowlege the financial support from China National Nuclear Corporation(CNNC) "Youth Talent Project" (JY21030603).

231

Abstracts - Oral

O8 - 16#8 Radioactive Waste and Current Radiological Issues

The improvement in the iodine capture property of silica-based
aerogel by introducing three-dimensionally ordered macroporous

Sen Chang1, Yongguo Li1, Jianli Chen1, Jie Yu1, Jianrong Hou1, Dangui Qiu1,
Qun Liu1,*
Department of Environment Engineering and Technology, China Institute for
Radiation Protection, China 1
Email: liuxingxin1@sina.com
Keywords: Dissolver off gas, Silver-silica aerogel, Three-dimensionally orderedmacroporous, Adsorption capacity, Gaseous iodine
Spent fuel reprocessing is the most critical link in the later stage of the nuclear
fuel cycle, and most processes such as cutting, dissolution, separation,collection and solidification will release gaseous iodine to varying degrees, ofwhich more than 90% of iodine will be volatilized into the dissolver off gas (DOG),
mainly mixed water vapor, nitrogen oxide and elemental iodine (I2), and will causeradiation hazards to the environment and organisms if directly discharged into
the atmosphere.
Iodine capture by porous materials is considered to be the most potential strategy
for purifying gaseous iodine. The reprocessing plants use silver-loaded silica gelor silver-loaded zeolite to capture the I2 in DOG, however, the single and wide
pore structure is difficult to meet the requirements of the large amount of 129I
capture. The emergence of the three-dimensionally ordered macroporous (3DOM)
structure provides a way to solve the above problems. It is urgent to carry out
structural design, synthesis methodology, establishment of the structureproperty relationship, and adsorption mechanism research based on integrated
material advantages.
Therefore, this work adopts a new process of sol-gel combined with colloidal
crystal template, and innovatively introduces 3DOM into silica aerogel system toform an inverse-opal-based structure with uniform macropore size, orderly
arrangement, and three-dimensional space interconnection.
3DOM can significantly reduce the diffusion resistance of gaseous iodine,enhance the "heat and mass transfer", and, make the active components
disperse through the "space barrier effect", inhibit the growth of the active
components, enhance the thermodynamic and dynamic characteristics of thereaction with gaseous iodine. Overall, a novel 3DOM composite aerogel with anadsorption capacity significantly higher than silver-loaded silica gel and silver-232

O8 - 16

#8 Radioactive Waste and Current Radiological Issues

loaded zeolite was successfully synthesized, providing new materials and newprocesses for the purification of radioactive iodine.
Acknowledgement
The authors are thankful for the support provided bythe China National Nuclear
Corporation (CNNC) Science Fund for Talented Young Scholars (No. JY20030603)
and the Basic Research Program of Shanxi Province (Free Exploration,
202203021211004).

233

Abstracts - Oral

O8 - 17#8 Radioactive Waste and Current Radiological Issues

Thermodynamic Analysis of Effect of Waste Resin Irradiation Gas
on Mineral Composition of Ordinary Portland Cement

HAN Xu1, GAO Chao, YAN Xiaojun, XI Yahui, QIN Xiang, LIU Enbang,
GUO Xiliang*
The Department of Radioactive Waste Management，
China Institute for Radiation Protection/China1
Email: Guoxl214@126.com
Keywords: spent resin irradiation gas, mineral composition, Gibbs free energy,
carbonation, cement degradation
Abstract: Recently, cement matrixes become a mainstream solution to address
the stabilization of radioactive spent resins. Generally, the cumulative dose ofspent resin cemented waste form will reach approximately 10^6 Gy over a disposal lifetime of 300 years. Due to the organic properties of spent resin, namely
producing gas if being irradiated, it inevitably affects the mineral phase of cement
waste forms. Therefore, the content and composition of the mineral phase could
be key factors determining the performance of cement waste forms. With that inmind, in this paper, the influence of the above factors on ordinary Portland cement (OPC) were explored especially. Specifically, based on the Gibbs energy
minimization principle, we explore the influence of three gases, namely, H2, CH4,
and CO2, generated from the irradiated waste resin towards the mineral phasecomposition and total volume belonging to the mineral phase of ordinary Portland
cement. Thereafter, the evolution of the mineral phase composition of ordinary
Portland cement over time is investigated based on the rate of irradiated gas
production. In this work, all the experiments are produced by GEMS software
based on the thermodynamic Gibbs free energy minimization method. Meanwhile,
our formulation for cement solidification of spent resin meets the standard of "Performance requirements for low and intermediate level radioactive waste formCemented waste form " (GB 14569.1-2011). From the experimental results, we
obtain several important findings. Firstly, only a small amount of H2 and CH4
reacts with the cement hydration products, which causes the dissolution of themineral phase with high Ca/Si, while the total volume of the mineral phase isalmost invariant. Secondly, in the prior phase, CO2 first reacts with the portlandite
in the cement hydration products. After the portlandite dissolute completely, thesilicate minerals start to dissolve, forming a system composed of abundant carbonate and a few clay minerals, which decreases the total volume of the mineral234

O8 - 17#8 Radioactive Waste and Current Radiological Issues

phase. Thirdly, during the predicted 160 years, the effects of spent resin irradiation gases on the cemented waste form are consistently dominated by CH4 andH2, while the performance of the cemented waste form is stable. However, therisk of rupture and explosion caused by the accumulation of H2 and CH4 in thewaste packaging containers and the risk of radioactive release due to carbonation of the cemented waste form by CO2 should not be overlooked. In summary,our research reveals the evolution law of how the waste resin irradiation gas
causes the degradation of cement waste forms. It could provide effective data
support for the safety evaluation of waste resin disposal by cement waste forms.

235

Abstracts - Poster List
No.

Topic

Title

Authors

P1-01

Operation test for themuon beam transportsystem of the μSR facility at RAON using a beta
radiation source

Kihong Pak1,
Jae Chang Kim1,
Jae Young Jeong1,
Yong Hyun Kim1,
Yong Kyun Kim1

P1-02

Validation of Nuclear
Data Libraries Using Radiation Shielding Benchmarks in SINBAD

Haelee Hyun1,
Do Heon Kim1,
Ser Gi Hong2

P1-03

Study of transformationof HTO in lettuce foratsmospheric release oftritiated water

Qiao Cao1,
Jianguo Li,
Shaofei Cao

P1-04

Assessment of Dose
Calculation Algorithms by Monte Carlo
Method(MCNP)

Changyeon Yoon1

P1-05

Analysis of Dose Assessment Methodology
by Boundary Voxel of
Irregular Shape Source
Based on Point Kernel
MethodMin Seong KIM1,
Ju Young KIM1,
Hyeok Jae KIM1,
Kwang Pyo KIM1

Evaluation of Neutron
effect for multi slit collimator in proton mini
beam radiation therapy

Chanil Jeon1,
Sangwoon Jeong1,
Youngyih Han2,
Sungkoo Cho2

P1-07

Research on 3D Radiation Field Calculation
Based on CAD Modeling

Zeyu Long,
Qingliang Feng,
Peitao Song,
Yuan Zhao,Hua Li

P1-08

Radiation Shielding andA c t i v a t i o n A n a ly s e sfor Neutron Activation
Measurement Facility at
ARTI, Jeongeup

Jounghwa Lee,Do Heon Kim,
Tae-Yung Song

P1-09

Monte Carlo optimization design of simulated
workplace neutron fields
for typical PWR workplaces

Faguo Chen1,
Hui Li1,
Deyuan Li1

P1-10

Improvement of integrated functions of thedose rate model for incident-free transportationof radioactive materials

Kiwon Jang,
Ara GoP1-11

Effects of Dose to Remaining Buildings fromthe Nuclear Power Plant
Decommissioning by
Shielding Materials andThickness

Hyung-Woo Seo,
Gang-Woo Ryu#1
P1-06 Radiation
Transport
andShielding236

Affiliation
1

Department of Nuclear Engineering, Hanyang University, Korea

1

Korea Atomic Energy Research Institute,Daejeon, Korea|2Department of Nuclear Engineering, Hanyang University, Seoul, Korea
1

Institute of Nuclear Environmental Science,
China Institute for Radiation Protection, China

1

KHNP Central Research Institute, Daejeon,
Korea

1

KyungHee University, Korea

1
Department of Health Sciences and Technology, SAIHST, Sungkyunkwan University,
Seoul 06355, Republic of Korea|2Departmentof Radiation Oncology, Samsung Medical
Center, Sungkyunkwan University School of
Medicine, Seoul, Republic of Korea

1

China Institute for Radiation Protection, China

1

Korea Atomic Energy Research Institute,Daejeon, Korea

1

China Institute for Radiation Protection, China

1

Department of Radiation Protection &Radioactive Waste Safety, Korea Institute of
Nuclear Safety, Korea

1

Decommissioning Technology Group, KHNP
CRI, Korea

Abstracts - Poster List
No.

Topic

Title

Authors#1
Radiation
P1-12 TransportandShieldingResearch and Verification of Radiation Shielding for a Non-destructive Testing Device of a
Coal Mine Belt

Jianyu Fu1,
Xiangming Xue1,
Jingming Zhan1

P2-01

Analysis of usefulness ofobesity risk factors using ROC curve for maleworkers

Baek Yun-Mi,
Kang Seo-gon,
Kim Jeong-in,
Sung Sook-hee

P2-02

D eve lopment of n-γ
mixed fields based on a
D-D neutron source forvalidation of material-filtered RPLGD for BNCT

Zixu Xu1,
Kazuma Aoki1,
Shingo Tamaki1,
Sachie Kusaka1,
Isao Murata1

Affiliation
1

China Research Institute of Radiation Protection, China

1

Korea Hydro & Nuclear Power Co., Ltd, Korea

1

Graduate School of Engineering, Osaka University, Japan

1
1

School of Nuclear Science and Technology, University of Science and Technology
of China, China| 2Department of Radiation
Oncology, First Affiliated Hospital of University of Science and Technology of China,
China|School of Nuclear Science and Technology, University of Science and Technology
of China, China|3Anhui Wisdom Technology
Company Limited, China

P2-03

Dosimetric comparisonof different techniquesof intraoperative radiotherapy for spinal metastasis based on MC
simulation

Junyi Liu ,Miao Qi1,
Ning Gao1,
Yikai Wu1,
Xi Pei3,
X.George Xu2

P2-04

Development of Computed Tomography-dedicated Mesh-Type Pediatric Phantom Library:Arms-removed andArms-raised Posture

Suhyeon Kim1,
Bangho SHIN1,
Chansoo CHOI2,
Hyeonil KIM1,
Haegin HAN1,
Sungho MOON1,
Gahee SON1,
Chan Hyeong KIM1

Image Reconstruction
with Limited-view-angle Projection Data for
BNCT-SPECT―Establishment of the Mock-up
System and Validation of
the Response Function―

Fan Lu,Kouki Okuda,
Haruka Inamoto,Shingo Tamaki,
Sachie Kusaka,
Isao Murata

P2-06

Establishing Local Exposure Scenarios for Local
Skin Dose Assessment
SystemSeung Beom
Goh1,
Han Cheol Yang2,
Chan Joong Kim1,
Jae Chang Kim1,
Kihong Pak1,
Yong Hyun Kim2,
Yong Kyun Kim1

1
Department of Nuclear Enginnering, Hanyang University, Korea|2Raymetrics, Co.,
Korea

P2-07

Numerical simulation for
calculating counting efficiencies considering thebiodistribution of I-131
in hand-held radiation
instrument

MinSeok Park1,
Jaeryong Yoo1

1
Department of Emergency Medical Research, Korea Institute of Radiological andMedical Sciences, Korea

P2-08

Haegin Han1,
Sungho Moon1,
Mesh-Type Phantom
Gahee Son1,
Posture Deformer: A
Bangho Shin1,
Real-Time GUI for PerHyeonil Kim1,
sonalized Phantom GenSuhyeon Kim1,
eration
Chan Hyeong Kim1

#2
Radiation
Dosimetry
P2-05

237

1

Hanyang University, Seoul, Korea|2J Crayton
Pruitt Family Department of Biomedical Engineering, University of Florida, Gainesville,
FL, USA

1

Department of Engineering, Osaka University, Japan1
Department of Nuclear Engineering, Hanyang University, Korea

Abstracts - Poster List
No.

Topic

Title

Authors

P2-09

Deep Learning for stable
Chromosome Analysis in
Bio-Dosimetry

LEE In Kyung,
KIM Jeong In,
BAEK Yoon Mi,
KANG Seung
Young

P2-10

Deep Learning for Unstable Chromosome Analysis in Bio-Dosimetry

LEE In Kyung,
HAN Jung Hun

P2-11

Sungho Moon1,
Enhancing Radiologist
Haegin Han1,
Pose Tracking in InterGahee Son1,
ventional Radiology: A
Bangho Shin1,
YOLOv7 and Multi-CamHyeonil Kim1,
era Approach for VirtualSuhyeon Kim1,
Dose Monitoring System Soon Young Song2,
Chan Hyeong Kim1

P2-12

Hyeonil KIM1,
Bangho SHIN1,
Chansoo CHOI2,
Suhyeon KIM1,
Haegin HAN1,
Sungho MOON1,
Gahee SON1,
Chan Hyeong KIM1

Development of GUIbased Monte Carlo
S i m u l a t i o n Co d e f o rExternal Exposure (McSEE): Implementation ofAdvanced Functions

P2-13

Bangho Shin1,
Chansoo Choi2,Evaluating Fetal RadiSuhyeon Kim1,
ation Exposure: DevelHyeonil Kim1,
opment and Impact ofHaegin Han1,
Pregnant-female MeshSungho Moon1,
type Reference ComGahee Son1,
putational Phantoms for
Yeon Soo Yeom3,
Accurate Dosimetry
Beom Sun Chung4,
Chan Hyeong Kim1

P2-14

Development of a prototype TL/OSL reader for
on-site application in alarge-scale radiologicalaccident

Hyoungtaek Kim1,
Chang-Young
Park2,
Sang In Kim3,
Min Chae Kim1,
Jungil Lee1,
Jang Lyul Kim1

P2-15

Evaluation of shielding
effect of personal dosimeter according toworking postures

Min Chae Kim1,
Hyoungtaek Kim1,
Sora Kim1,
Yoomi Choi1,
Jiyoon Kim1,
ByungIl Min1,
Kyungsuk Suh1,
Jungil Lee1,
Haegin Han2,
Chan Hyeong Kim2

P2-16

Dose Evaluation of High-energy Photons Leaking from Kindai Reactor

Shintaro KIMURA1,
Hirokuni
YAMANISHI2#2
Radiation
Dosimetry

238

Affiliation
1

RHI, Korea Hydro & Nuclear Power Co., LTD,
Korea

1

RHI, Korea Hydro & Nuclear Power Co., LTD,
Korea

Hanyang University, Korea|2Hanyang University Hospital, Korea

1

1

Department of Nuclear Engineering, Hanyang University, Seoul, Korea| 2J Crayton
Pruitt Family Department of Biomedical Engineering, University of Florida, Gainesville,
FL, USA

1

Department of Nuclear Engineering, Hanyang University, Seoul, Korea| 2J Crayton
Pruitt Family Department of Biomedical Engineering, University of Florida, Gainesville,
FL, USA|3Department of Radiation Convergence Engineering, Yonsei University, Wonju, Korea|4Department of Anatomy, Yonsei
University Wonju College of Medicine, Wonju, Korea

1

Radiation Safety Management Division, Korea Atomic Energy Research Institute, Daejeon, Korea|2Dept. of Physics, Gyeongsnag
National University, Jinju, Korea|3ARIM Science, Daejeon, Korea

1

Korea Atomic Energy Research Institute(KAERI), Republic of Korea|2Department of
Nuclear Engineering, Hanyang University,
Republic of Korea

Grad.Sch.Kindai Univ, Japan|2AERI,Kindai
Univ, Japan1

Abstracts - Poster List
No.

Topic

Title

Authors

Affiliation
1

P2-17

Soo Kyung Jeong ,Su Jung Oh1,
Comparing the Perfor- Yeong-Rok Kang1,
mance of an Automatic
Hyo Jin Kiim1,
S y st e m w i t h M a n u a l Chang Geun Lee1,
Scoring in the Dicentric Song-Hyun Kim2,
Chromosome Assay
Seungsoo Jang3,
Min Ho Jeong4,
Wol Soon Jo1

P2-18

Neutron Dose Inside a
50 MeV Cyclotron Vault

Ahn Yunjin1,
Kim Sangrok1

1

P2-19

Uncertainty Evaluation ofcalibration of a Beta-ray
Irradiation System using
the Monte-Carlo Method

Moonhyung Cho,
Hyunjin Kim

1

P2-20

I m p ro v e m e n t o f t h e
Performance of a Whole
Body Counter Using an
Artificial Neural Network

Moonhyung Cho1,
Hyunjin Kim1

1

P2-21

Research on mRNA asa Biomarker of Mixed
Neutron and γ -ray Irradiation

Yuan Yayi 1,
Chai Dongjing1,
Dong Juancong1,
Jiao Cheng1,
Dang Xuhong 1

P2-22

Establishment and application of the biokinetic
model of plutonium livercells

Shuang-Shuang
CUI1,
Juan-Cong DONG1,
Jiao CHENG1,
Hong-Yan LIU1,
You-Chen LI1

Institute of Radiology and Environmental
Medicine, China Institute for Radiation Protection, China

P3-01

Ju Hyung Kim1,
Development of a MethYong Hyun Kim2,
odology to Convert theKi Hong Park2,
Energy Spectra to SpecSeung Beom Goh2,
tra for Channel
Yong Kyun Kim2

1
Department of Nuclear Engineering, Hanyang University, Seoul, South Korea |2RayMetrics, Seoul, South Korea

P3-02

Correction of changes
in Photopeak Position
and FWHM according toTemperature Changes ofCsI(Tl) Detector

ChanJoong Kim1,
YongHyun Kim2,
HanCheol Yang2,
JunehyungBernaski1,
JooHyung Kim1,
YongKyun Kim1

Evaluation of Output
Accuracy and Precisionof 3D-Printed Scintillator(RMPS470)

JunehyungBernaski1,
Jae Chang Kim1,
Yong Hyun Kim2,
Han Cheol Yang2,
Yong Kyun Kim1

#2
Radiation
Dosimetry#3
Radiation
Detectionand Sensor
P3-03
Technology

P3-04

Xuan Hou1,
Hironobu Yanagie2,
Improvement and evalTakehisa Matsu
uation of anti-tumor efkawa3,
fect by neutron capture
Daibing Yang1,
therapy using a new elMasashi Yanagawa4,
ement — rhodium based
Kenji Shimazoe1,
on beta ray
Hiroyuki
Takahashi2239

1

Dongnam Institute of Radiological & Medical
Sciences, Korea|2KEPCO International Nuclear Graduate School (KINGS), Korea|3Divisionof Advanced Nuclear Engineering, POSTECH,
Korea|4Dong-A University College of Medicine, Korea

Radiation Safety section,Korea Institute ofRadiological and Medical Sciences, Korea
Central Research Institute, Korea Hydro &Nuclear Power Co., Ltd., Korea

Central Research Institute, Korea Hydro &Nuclear Power Co., Ltd., Korea

1
Department of Radiation Medicine and Environmental Medicine, China Institute for
Radiation Protection, China

1

1

Department of Nuclear Engineering of Hanyang University, South Korea|2Raymetrics,
South Korea

1

Department of Nuclear Engineering, Hanyang University, Korea|2Raymetrics, Seoul,
Korea

1

Department of Bioengineering, Schoolof Engineering, The University of Tokyo,
Japan|2Institute of Engineering Innovation,
School of Engineering, The University of Tokyo, Japan|3Department of Epidemiology andEnvironmental Health, Juntendo University,
Japan| 4Veterinary Medical Center, Obihiro University of Agriculture and Veterinary
Medicine, JapanAbstracts - Poster List
No.

Topic

Title

Authors

Affiliation
1

P3-05

Evaluation of Properties
of Plastic Scintillators
Fabricated by 3D Printing

Han Cheol Yang ,Tae Hoon Kim2,
Yong Hyun Kim2,
Yong Kyun Kim1

P3-06

Accelerated Monte Carlo-Based Gamma-Ray
Spectrometer Modelling
for Efficient Dataset
Creation in Radiation
Detection and Sensor
Technology

Young-su Kim1,
Jeen-hyeng
Kwon2,
Jiseok Kim1

P3-07

Surface contamination
meters and monitors
based on plastic scintillator–WLS–SiPM technology produced on SPC‘’Doza”

Pavel Buzhan1,
Aleksandr
Karakash1

P3-08

Implementation of Digital Pulse Processing for
Neutron Measurement
Using Recursive Algorithms

Jae Chang Kim1,
Junehyung Lee
Bernaski1,
Yong Kyun Kim1

P3-09

Coded Aperture Gamma
Camera Prototype Basedon CeBr3 Single Crystal
Position-Sensitive andSpectrometric Detector

Yury Teverovskiy1,
Alain Iltis2,
Ghislain Zeufack2#3
Radiation
P3-10 Detectionand Sensor
Technology

A study on methods tominimize pulse pile-up
and dead time errors inhigh radiation fields

Sujung Min,
Bumkyoung Seo,
Sangbum HongP3-11

Evaluation of radiation
detection properties ofliquid scintillators basedon modified scintillationmaterial

Sujung Min,
Sangbum Hong,
Soon Park,
Changhyun Roh,
Bumkyoung Seo

P3-12

Joonhyuk LEE1,
Changyu KO1,
Development of miniaMinki CHAE1,
ture gamma-ray imager
Dukwoo KIM1,
using 3D printed tungSeungho LEE1,
sten coded aperture
Shinhye KANG1,
Youngkwang KIM1,
Manhee JEONG1

P3-13

Enhancing Proton Therapy
Range Verification: The
Multi-Slit Prompt-Gamma Camera for Improved
Conformal Dose Delivery

Youngmo Ku1,
Sehoon Choi1,
Jaeho Cho1,
Sehyun Jang1,
Jong Hwi Jeong2,
Sung Hun Kim2,
Sungkoo Cho3,
Chan Hyeong Kim1

P3-14

Accelerating Cargo Screening: Development and
Evaluation of a Hybrid
Gamma Imaging System
for Rapid Detection of
Nuclear and Radioactive
Materials

Junyoung Lee,
Goeun Lee,
Taehyeon Eom,
Youngmo Ku,
Chan Hyeong KIM

1Department of Nuclear Engineering, Hanyang University, Korea|2Raymetrics, Korea

1240

Korea Atomic Energy Research Institute,Daejeon 34057, Republic of Korea| 2 Department of Nuclear Engineering, Hanyang
University, Seoul 04763, Republic of Korea|Korea Atomic Energy Research Institute,Daejeon 34057, Republic of Korea

1

Scientific Production Company "Doza", Russian Federation

1

Department of Nuclear Engineering, Schoolof Hanyang University, Korea

1

Scientific Production Company "Doza", Russian Federation

1
Department of decommissioning technology
development, Korea Atomic Energy ResearchInstitute, Korea

1

Department of decommissioning technology
development, Korea Atomic Energy ResearchInstitute, Korea

1

Jeju National University, Korea

1

Depaprtment of Nuclear Engineering, Hanyang University, Korea| 2Proton Therapy
Center, National Cancer Center, Korea|3Department of Radiation Oncology, Samsung
Medical Center, Sungkyunkwan University
School of Medicine, Korea

1

Department of nuclear engineering, Hanyang University, Korea

Abstracts - Poster List
No.

Topic

Title

AuthorsP3-15

Addressing Temperature
Variations in Outdoor
Gamma Imaging: A
Novel Regression-BasedPMT Gain Optimization
TechniqueGoeun Lee1,Junyoung
Lee1,Youngmo
Ku1,Chan Hyeong
KIM1

P3-16

Development of CZTbased Portable Gamma
Imaging System for Korea Customs Service

Jae Hyeon
Kim1,Jin Hyung
Park1,Byeong-no
Lee1,Junyoung
Lee2,Kyung Min
Oh1

P3-17

Fe a s i b i l i t y S t u d y o n
Pulse Shape Discrimination of Fast/Thermal
Neutrons, and Gamma
ray with a Single Detector

Chaehun
Lee,Seonkwang
Yoon,Byung-Hee
Won,Seong-Kyu
Ahn,Sang-bum
Hong

P3-18

Application of Optical Fiber Beam Loss Monitor
in SSRF

Dong Zhang1,Ming
Zeng2,Jianbing
Zhang1,Jiaqiang
Xu1

1
Department of Radiation Protection andTechnical Safety, Shanghai Synchrotron Radiation Facility, Shanghai Advanced ResearchInstitute, China|2Department of Engineering
Physics, Tsinghua University, China

P3-19

Time-of-Flight Neutron
B ra g g Ed g e I m a g i n g
Using a Compact Accelerator Neutron Source

Mahdi Bakhtiari1,Hirotaka
Sato2,Nam Suk
Jung3,Wooyong
Um1,Hee Seock
Lee3

1
Division of Advanced Nuclear Engineering,
Pohang University of Science and Technology, Korea|2Faculty of Engineering, Hokkaido
University, Japan|3Poahng Accelerator Laboratory, Pohang University of Science andTechnology, Korea#3
Radiation
Detectionand Sensor
P3-20
Technology

Affiliation
1

Department of nuclear engineering, Hanyang University, Korea

1

Advanced Radiation Technology Institute,
Korea Atomic Energy Research Institute,
Korea|2Department of Nuclear Engineering
Hanyang University, Korea

1

Korea Atomic Energy Research Institute,
Korea

A machine learningbased peak area calculation algorithm in low
background measurement

Peng Zhang

Key Laboratory of Particle and Radiation
Imaging (Ministry of Education) and Department of Engineering Physics, Tsinghua University, Beĳing 100084, China

P3-21

Experimental verificationof real-time gamma-ray
energy spectrum anddose monitor up to 3
MeV

Nikolaos Voulgaris1,Hayato
Sumi1,Sachie
Kusaka1,Tamaki
Shingo1,Isao Murata1

1
Division of Sustainable Energy and Environmental Engineering, Graduate School ofEngineering, Osaka University, Osaka, Japan

P3-22

Youngjoo Park1,Inbum Lee1,Kangwoo Jeon2,DonRemote Radiation Detection Imaging System ghui Seo1,Kyunghoon Cho2,Jaewon
Mounted on a Drone
Jeong2,Hakjae
Lee2,Kisung Lee1

P3-23

Development of a Coded
Aperture Collimator for
Drone-Based Radiation
Imaging System

Inbum
Lee1,Youngjoo
Park1,Kangwoo
Jeon2,Donghui
Seo1,Kyunghoon
Cho2,Jaewon
Jeong2,Hakjae
Lee2,Kisung Lee1

P3-24

Simulation of boroncoated honeycomb tube
neutron detector withhigh detection efficiency

JianjinZhou1,Hengguan Yi1,Sanqiang
Xia1,Hengchun
Zhao1,Qinjian Cao1

1

1

241

1

Department of Biomedical Engineering,
School of Korea University, Korea|2ARALE
Laboratory Co., Ltd, Korea

1

Department of Biomedical Engineering, Korea University, Seoul, Korea2ARALE Laboratory Co., Ltd, Korea

1

China Institute for Radiation Protection,
China|2Shanxi Provincial Key Laboratory for
Radiation Safety and Protection, China|3CNNC Key Laboratory for Radiation Protection
Technology, China

Abstracts - Poster List
No.

Topic

P3-25

Title

Authors

Affiliation

Characteristic Determination of 1,800 cc Ion
Chamber and Radiation
Measurement for Human Body Scanner

Yong Ho Jin1,
Min Woo Kwak1,
Hyun Woo Kim1,
Kwang Pyo Kim1

1

Kyung Hee University, Yongin, Korea

1
Advanced Fuel Cycle Technology Development Division, KAERI, Korea| 2Advanced
Fuel Cycle Technology Development Division, KAERI, Korea| 3Advanced Fuel Cycle
Technology Development Division, KAERI,
Korea|4Department of Quantum System Engineering, JNU, Korea|5Advanced Fuel Cycle
Technology Development Division, KAERI,
Korea

P3-26

Neutron multiplicationa n d p ro m p t n e u t ro nangular distributions ofdiverse nuclear material
samples

Seonkwang Yoon1,
Chaehun Lee2,
Young-Su Kim2,
Hee Seo2,
Ho-Dong Kim2

P3-27

Development of a New
Low-Energy Neutron
Spectrometer Using a
Position-Sensitive Proportional Counter -Examination of Response
Functions Including
Shield-Yu Fujiwara1,
Masahiro Hori1,
Shingo Tamaki1,
Sachie Kusaka1,
Fuminobu Sato1,
Isao Murata1

P3-28

The performance analysis of the laundry monitor

Jing Zhang1,
Yi Ren1,
Xiangyang Du1,
Jianfei Wang2,
Xirong Guo1

#3
Radiation
P3-29 Detectionand Sensor
Technology

Investigation of a Statistical Spatial Analysis
Methodology for Measurement-based Target
Point’s Spatial Dose Rate
Estimation.

Ji Woo Kim,
Hee Won Lee,Min Seong Kim,
Kwang Pyo Kim

P3-30

Development of Online
Directional Dose Equivalent Rate Monitoring
Systemyingjing wei1,
liye liu2,
zhifang wu1

Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing, China|2China Institute for radiation protection,
Taiyuan, China

P3-31

Design and simulation
verification of a coded
aperture gamma camera
with different verticaland horizontal fields ofview

Yan Li1,
Hengguan Yi1,
Sanqiang Xia1,
Chongyang
Wang1,
Xiaolong Wang1

1
Shanxi Provincial Key Laboratory for Radiation Safety and Protection, CNNC Key Laboratory for Radiation Protection Technology,
China Institute of radiation protection,China

P3-32

Application of new UAV
radiation aerial survey
technology in nuclear
emergency

mei feng1

P3-33

Monte Carlo Simulation
for An Underground Radioactivity Measurement

Chen Jikai1

P3-34

3D radiation mapping:3D scene and gamma
radiation dose rate measurement data fusion

Hui LI1,
Qing FAN1,
Hua LI1,
Faguo CHEN1,
Liye LIU1,
Sanqiang XIA2,
Yuan ZHAO3

P3-35

Re s e a rc h sta t u s a n ddevelopment of neutron
detector

1

Osaka University, Japan

1

China Institute for Radiation Protection,
China|2Shanxi Zhongfu Nuclear Instruments
Co.Ltd, China

1

Department of Nuclear Engineering, Kyung
Hee University, Korea

1

1

Department of Health Physics, China Institute for Radiation Protection, China

1

Department of Engineering Physics, Tsinghua University, China

1
China Institute for Radiation Protection,
Taiyuan, China|2Shanxi Provincial Key Laboratory for Radiation Safety and Protection,
Taiyuan, China|3Key Laboratory of Radiation
Protection Technology of China National Nuclear Corporation, Taiyuan, China
11hengchun zhao

242

China Institute for Radiation Protection,
China|2Shanxi Key Laboratory for Radiation
Safety and Protection, China

Abstracts - Poster List
No.

Topic

P4-01

Title

Authors

Natural radioactivity inbottled water and associated age-dependent
dose and lifetime risk
assessment in Korea

Ji Eun Lee,
Hyo Jin Kim,
Yong Uk Kye,
Hee Jin Jang,
Yoon Ha Kim,
Wol Soon Jo,
Chang Geun Lee,
Jeung Kee Kim,
Yeong-Rok Kang

Affiliation

1

Dongnam Institute of Radiological & Medical
Sciences, Seoul, Korea

1

P4-02

Optimization of thedetection system forreal-time monitoring ofmarine radioactivity

Mee Jang1,
Hyun Chul Kim2,
Min Sun Lee3,
Young Hak Kim4,
Jae Hyeon Seo5,
Jong Myoung Lim6

P4-03

A study on the methodof measuring environmental level 137Cs radioactivity in seawaterDong Hye Heo1,
Eun Hee Kwon2,
S.H. Hwang,B.C.
Kim,
K.B. Lee,J.M. Lee

One-Class SVM Based
Anomaly Detection for
LSTM-Based Radiation
Prediction

sihyun lee1,
raehyun lee

P4-05

An Approach to Calculation of Individual Doses
on Explosion Accident
Scenarios during Decommissioning of Nuclear Power Plants

Kwangho Jo1,
Hyu-Chang Choi,
Jung Hyejin,
Ryu Gangwoo

P4-06

LSC method for determination of Total-Uranium and Uranium Isotopes in Liquid Effluentsfrom nuclear fuel cycle
facilities

Yang Youkun1,
Bao li1,
Guo chen1,
Yang Hailan1

P4-07

Main Performance Testof Gaseous Effluent
Sampling System

Li Qiao1,
Yi Zhang,
Haiyu Hong,Jianfei Wang

P4-08

Gamma nuclides discrimination technology
based on fully connectedneural network

Xin Liu1,
Ri Zhao1,
Na Liu1,
Jing Zhang1,
Runcheng Liang1,
Zhaoxing Liu1

P4-09

Research on computational efficiency optimization algorithm basedon source weight discrete

Wei Jiafu1, Li Hua1,
Zhao Yuan1,
Jin Chenghe1,
Song Peitao1,
Cao Qinjian1,
Long Zeyu1,
Wang Yu1#4
EnvironP4-04 mental
Radiation
Measurement andAssessment

243

Environmental Radioactivity Assessment
Team, Korea Atomic Energy Research Institute, Korea|2Environmental Radioactivity Assessment Team, Korea Atomic Energy Research Institute, Korea|3Environmental Radioactivity Assessment Team, Korea Atomic
Energy Research Institute, Korea|4Environmental Radioactivity Assessment Team,
Korea Atomic Energy Research Institute,
Korea|5Environmental Radioactivity Assessment Team, Korea Atomic Energy ResearchInstitute, Korea|6Environmental Radioactivity
Assessment Team, Korea Atomic Energy
Research Institute, Korea
1

Ionizing Radiation Metrolgy Group, Divisionof Chemical and Biological Metrology, Korea
Research Institute of Standards and Science(KRISS), Korea| 2University of Science andTechnology (UST), korea
1

Department of Electronic Engineering, Korea

1
Decommissioning Technology Group, KHNP
CRI, KOREA

1

China Institute for Radiation Protection, Taiyuan, China

1

China Institute for Radiation Protection, Taiyuan Shanxi, China

1

China Institute for Radiation Protection, China

1

China Institute for Radiation Protection, China

Abstracts - Poster List
No.

Topic

Title

Authors

Distribution of Naturally
Occurring Radioactive
Materials in Groundwater of Jeju Island

Sohyeon Lim1,
Chung-Hun Han1,
Hee-Jung Im1

P4-11

Particle Size Distributionof Radioactive Aerosols
in Workplace

Feng Zhang1,
Pengxiang Li1,
Ruijun Wang1,
Jinhao Li1,Yu Li1,
Yongheng Guo1,
Yong Li1

P4-12

Pretreatment Method
a n d A s h F re s h Ra t i o
Calculation of Biological Sample Analysis in
Radiation Environment
Monitoring

Jinhao Li1,
Feng Zhang1,
Ruijun Wang1,
Qinnan Song1,
Yongheng Guo1,
Yong Li1

P4-13

Cooperation on Radia t i o n M e a s u re m e n t s
for Cross-Border Accidents (2) Application ofmachine learning conversion method using
monitoring data from
Fukushima as teachingdata

Miyuki Sasaki1,
Kotaro Ochi1,
Yukihisa Sanada1,
Sungyeop Joung2,
Yoomi Choi2,
Eunjoong Lee2,
Wanook Ji2,
Young-Yong Ji2#4
Environmental
P4-14 Radiation
Measurement andAssessment

Cooperation on Radia t i o n M e a s u re m e n t s
for Cross-Border Accidents (1) A CZT detector-based airborne survey analysis in high dose
rate environment

Sungyeop Joung1,
Young-Yong Ji1,
Yoomi Choi1,
Eunjoong Lee1,
Wanook Ji1,
Miyuki Sasaki2,
Kotaro Ochi2,
Yukihisa Sanada2

Environmental Safety Technology Research
Division, Korea Atomic Energy ResearchInstitute, Korea|2Environmental Monitoring
Group, Japan Atomic Energy Agency, Japan

P4-15

Feasibility Study of Directly Measuring Radon
Concentration in Air Using TPC

Qiuli Zhang1

1
Department of Engineering Physics, Tsinghua University, China

Feifei Wu1,
Yan Wang1,
Jing KANG1,
Bing Lian1

1

P4-16

Life cycle greenhouseg a s e m i s s i o n s f ro m
nuclear power chain in
China

Meng Wang1,
Bing Lian1,
Yan Wang1,
Jing Kang1

1

P4-17

Study on Optimal Control
of Liquid Effluent Emission from Small Nuclear
Power Reactor

P4-18

Determination of 226Ra
in drinking water inSouthern China by Liquid Scintillation Counting

CHEN GUO,HAILAN YANG,YOUKUN YANG,
LI BAO,RUIJUN WANG

Combined transuranicnuclide analysis.of aerosol samplese

Ke Du1,Peng Xiang Li1,
Xu Yuan Ma,
Yu Hu Han,
Feng Zhang,
Li Bao,Rui Jun Wang

P4-10

P4-19

244

Affiliation
1

Department of Chemistry, Jeju National
University, China

1

China Institute for Radiation Protection, China

1

China Institute for Radiation Protection, China

1

Japan Atomic Energy Agency, Fukushima,
Japan|2Korea Atomic Energy Research Institute, Daejeon, Korea

1

China Institute for Radiation Protection,
Key Laboratory of Radiation Environment &Health of the Ministry of Ecology and Environment, Taiyuan, 030006, China
China Institute for Radiation Protection,Key
Laboratory of Radiation Environment &Health of the Ministry of Ecology and Environment, Taiyuan, 030006，China

1

China Institute for Radiation Protection, China

1

China Institute for Radiation Protection, China

Abstracts - Poster List
No.

Topic

Title

Authors

P4-20

The Content of Tritium
and Carbon-14 in Env i ro n m e n t a l M e d i u maround Shidaowan High
Temperature Gas Cooled
Reactor NPPQin Nan SONG1

P4-21

Assessment of Uncertainty in Collecting
Radioactive Soil Samples by Sampling andAnalytical Quality Control（SAX）

Ruijun Wang,
Qinnan Song,Li Bao,Jinhao Li,
Xvjia Hou,
Haiyang Wu,
Qiqi Shi

P4-22

Evaporation behaviorof polonium-210 fromlead-bismuth eutectic at
low temperature

Huiyun Xue1,
Yong Guo2,
Ping Xu1,
Guobing Yu3,
Yujia Kang1,
Zhi Chen1

P4-23

Wind tunnel study ofeffects of thermal stratification on turbulent
flow and diffusion over athree-dimensional hill

Yun-peng LI1,
Ruo-jie LI1,
Yan-hui Pan1,
Si-bo Lyu1,
Zhan-jie Guo1,
Jun-fang Zhang1

1
China Institute for Radiation Protection, China

Background Spectrum
S i m u l a t i o n D eve lo p ment and Verification for
In-Situ Measurement
Technology Application

Byungchae Lee1,
Jihyun Yu1,
Jonghoa Kim1,
Jangsoo Suh1,
Sangbum Hong2

1
R&D Institute, Sae-An Enertech Corp, Korea|2Korea Atomic Energy Research Institute,
Korea

#4
Environmental
Radiation
Measurement andP4-24 Assessment

Affiliation

1

China Institute for Radiation Protection, China

1

Key Laboratory of Radiation Environment &Health of the Ministry of Ecology and Environment, China Institute for radiation proctection, China

1

University of Science and Technology ofChina, Anhui Province, China| 2China Nuclear Power Technology Research Institute,
518000 Guangdong, China|3Anhui Radiation
Environmental Supervision Station, Anhui
Province, China

P4-25

Sensitivity Analysis of Seong Yeon LEE1,
Distribution Coefficient Seong Hun JEON1,
by Nuclide According to Min Seong KIM1,
Kwang Pyo KIM1
Landfill of NORM Waste

P4-26

A Study on Radiological Characterization for
Alpha Source Terms in
Decommissioning Nuclear Power Plants

Seong Jun Kim1,
Tae Young Kong 1

1

Eunjoong Lee1,
Byoungil Jeon1,
Young-Yong Ji1

1

P4-27

Deep learning-based
spectrum-dose conversion model for KAERIMARK-M detection system

P4-28

Study of the radiation
effects of radioactive
residues in soil at a decommissioning site

ZhiXiang Yu

245

1

Department of Nuclear Engineering, Kyung
Hee University, Korea

Department of Nuclear Engineering, Schoolof Chosun University, Korea

Environmental Safety Technology Research
Division, Korea Atomic Energy Research Institute, Korea

1

China Institute for Radiation Protection, China

Abstracts - Poster List
No.

Topic

Title

Authors

P4-29

Analysis of Carbon-14
Radioactive Effluents
Released from Korean
Nuclear Power Plants
and Its Resulting Public
Dose

Hwa Pyoung KIM1,
Tae Young KONG1

P4-30

Specification for monitoring of uranium aerosol
concentration in the air
at workplaces of uranium processing and fuelfabrication facilities

Baoli WU1,
Xiaona GU1,
Xiangming XUE1,Kai YANG1

Effects of Scenarios andSource Types on Dosesfor the Reuse Buildings
from the Decommissioning of Nuclear Facilities

Hyung-Woo Seo,
Gang-Woo Ryu

Decommissioning Technology Group, KHNP
CRI, Korea

P4-32

Review of Dose Estimation Considering Decay
Correction for the Decommissioning of Nuclear Power Plants

Hyung-Woo Seo,
Ji-Hwan Yu,
Gi-Lim Kim

1
Decommissioning Technology Group, KHNP
CRI, Korea

P4-33

Re l a t e d re s e a rc h o necological environmental impact assessment
around nuclear fuel reprocessing plant

Gong Wenjing1,
Cao Shaofei1,
Li Jianguo1,
Wang xinni1

P5-01

Impact of vendor-agAhyeong Lee1,
nostic deep learnChulkyun Ahn1,
ing-based CT denoising
Hyeung Woo Nam2,technique on patient’s
Chang Won Kim3,
radiation dose reduction
Kwang Pyo Kim2,
in a multi-vendor CT
Jong Hyo Kim1
clinical environment

P5-02

Rev i ew o n Ra d i a t i o nProtection Program
during Decommissioningof NPP

Hye Jin Jung1,
Hyu Chang Choi,
Kwang Ho Jo,
Hyung Woo Seo,
Jin Won Son

Current Status Analysis
to Optimize Shielding
Design Criteria for Radiotherapy Facilities using Linear Accelerators

Sang Hyun Park1,
Jong Hoon Park1,
Beom Suk Ko1,
Sung Hoon Ko1

P5-04

Review of the radiationdose of aircraft crew due
to cosmic radiation

Byungmin Lee,
Jaeryong YOO

P5-05

A Review on Factors
Necessary for Estimating
Worker Exposure Dosefor Decommissioning
Nuclear Power Plants

Gi-Lim Kim1,
Hyung-Woo Seo1,
Minchul Kim1,
Young-il Na1,
Jihwan Yu1#4
Environmental
Radiation
P4-31 Measurement andAssessment#5Radiological
Risk
P5-03 Management

246

Affiliation

1

Chosun university Department of Nuclear
Engineering, Korea

1

China Institute for Radiation Protection, China

11China Institute for Radiation Protection,
Shanxi, Taiyuan, China

1

Department of Applied Bioengineering,
Graduate School of Convergence Scienceand Technology, Seoul National University,
Rep. of Korea|2Department of Nuclear Engineering, Kyung Hee University, Rep. of Korea|3ClariPi Research, Seoul, Rep. of Korea

1

Decommissioning Technology Group, KHNP
CRI, Korea

1

Korea Institute of Nuclear Safety, Korea

1

Korea Institute of Radiological and Medical
Sciences, Seoul, Korea

1

KHNP CRI, Korea

Abstracts - Poster List
No.

Topic

Title

Authors

Co m p a r i s o n o f D o s eAssessment of Off-link
Public According to Radioactive Waste Transportation by Computer
Code

Minwoo Kwak1,
Geonwoo Son1,
Hyeokjae Kim1,
Kwangpyo Kim1

Assessment of Accident
Probability for Overland
Transportation of Spent
Nuclear Fuel in Korea

Jooyub Lee1,
Jaehyo Kim1,
Geehyun Kim1,
Kiwon Jang2,
ARA GO2,
Han-Ki Jang3

P5-08

Evaluation of Preliminary
Risks Derived for theD e co m m i s s i o n i n g o f
Nuclear Facilities

Hyung-Woo Seo,
Gi-Lim Kim

P6-01

Development of Quantitative Assessment
System for Evaluationof Training Competencein Virtual Reality-based
LINAC QA Training

Jongeun Kim ,Wonhyeong Lee2,
Yoonsun Chung2

P6-02

Comparison of Radiation
Regulatory Systems in
Selected Countries :Focusing on Differences
and Adaptations

Euna Lee1,
Yoonsun Chung1

Analysis of Radiation
Monitoring System for
Imported Cargo in Korea,
the United States, theUnited Kingdom, andJapan

Choi Wo Suk1,
Kong Tae Young1

P6-04

Student Training Program for Radiation Safety in Hanyang University:
HY-Radiedu programand Radiation Safety
technology conference

Euna Lee1,
Wonhyeong Lee1,
Euidam Kim1,
Junik Cho1,
Jongeun Kim1,
Chae-eon Kim1,
Jiwon Choi1,
Yoonsun Chung1

P7-01

Review of Establishing
EPZ through GIS tools

Bongseok Kim1

D eve lo p m e n t o f G U I
Program for Dose Assessment for Skin Contamination in Radiation
Accident

Lee Seung Je1,
Kim Han Sung1,
Yoo Jaeryong1

Derivation of Considerations for Developmentof Victim Management
Procedure in Radiation
Emergency

Kihoon KIM,
Jaeryong Yoo

P5-06

#5
Radiological
Risk
P5-07
Management#6
Education,
Training
and Policy
P6-03 in Radiation
Safety#7
RadiologicalP7-02
Emergency
Planning
and Preparedness
P7-03

Affiliation

1

Department of nuclear engineering, Kyunghee University, Korea

1

Department of Nuclear Engineering, Seoul
National University, Korea|2Department ofRadiation Protection and Radioactive Waste
Safety, Korea Institute of Nuclear Safety,
Korea|3Department of Future Strategic Business Bureau, Korea Association for Radiation
Application, Korea

1
Decommissioning Technology Group, KHNP
CRI, Korea

1

1

Department of Nuclear Engineering, Hanyang University, Seoul, Korea|Proton Therapy Center, National Cancer Center, Goyang,
Korea|2Department of Nuclear Engineering,
Hanyang University, Seoul, Korea

1

Department of Nuclear Engineering, Hanyang University, Seoul, Korea

1

Department of Chosun University, Korea

1
Department of Nuclear Engineering, Hanyang University, Seoul, Korea

1

247

Nuclear Emergency Preparedness Team,
Korea Atomic Energy Research Institue, Korea
1National Radiation Emergency Medical Center, Korea Institute of Radiological and Medical Sciences (KIRAMS), Seoul, Korea

1

Korea Institute of Radiological & Medical
Sciences, Seoul, Korea

Abstracts - Poster List
No.

Topic

Title

Authors

P7-04

Development of a Screening Method for Contamination in Urine Samples
during Radiation Emergencies

Seokwon Yoon 1,
Jaeryong Yoo1

P7-05

Radiation exposure dose
reduction in radiation
emergency medicine

Jaeryong Yoo,
MinSeok Park,
Kihoon Kim,
Byungmin Lee

P7-06

Analysis of MountainValley Breeze Morphology Affecting the Diffusion and Distribution of
Pollutants at an Inland
Nuclear Facility

Huang Sha1,
Zhao Duoxin1,
Niu Yanjing1,
Zhang Yadong1

P7-07

Comparison of diffusion
simulation results of
JRODOS and NACADOS
at a coastal plant site in
China

Zhao Duoxin1,
Huang Sha1,
Lv Minghua1,
Chen Longquan1

P7-08

Emergency Action Levels Development Studyfor Research Reactors

ilje Cho,
Wiho Ha,
Chanki Lee,
Wontae Hwang,
Hyunki Kim

Ra p i d d e t e r m i n a t i o nof Pu, Am and Cm inurine sample by tandem
quadrupole ICP-MS

Yang Wu1,
Maoyi Luo1,
Xiongxin Dai1,
Ni Yuan1,
Hongmei Guo1,
Jingjing He1

P7-10

Effects of Neutron, γ-Ray, and Neutron–γ-Ray Mixed Radiations
on Chromosomal Aberrations in Peripheral-Blood Lymphocytes

Zhang Zhongxin1,
Bai Jiamei1,
Meng Qianqian1,
Dang Xuhong1,
Liu Hongyan1,
Yuan Yayi1

P7-11

Single-column Chromatography Separation for
Simultaneous Determination of Pu, Np, U, Th
in Small-volume Urinefor Emergency Radioassays

Maoyi LUO,
Xiongxin DAI,
Yang WU,
Pengyu WANG,
Jingjing HE,
Yuan NI

P7-12

Establishment of the
Dose-response Calibration Curve for Semi-automatic Analysis of Dicentric Chromosome

Qianqian Meng,
Zhongxin
Zhang,Yue Ren,
Xiaozhen Li,
Yayi Yuan,
Jingjie Wang,
Ruifeng Zhang

P7-13

Role of TLR4/NF-κ
B signaling pathway inh y d ro g e n - r i c h w a te r
improving acute radiation-induced brain injury
in rats

mengya liu1,
xiaoman chen2,
yong wang2,
hui yuan1,
huan liu1,
jianguo li1,
xiujun qin1

#7
P7-09 Radiological
Emergency
Planning
and Preparedness248

Affiliation
1

Korea Institute of Radiological and Medical
Sciences, Korea

1

Korea Institute of Radiological and Medical
Sciences, Seoul, South Korea

1

China Institute for Radiation Protection, Taiyuan, China

1

China Institute for Radiation Protection, Taiyuan, China

1

Nuclear Emergency Prepardness Section,
Korea Atomic Energy Research Institue, Korea

1

China Institute for Radiation Protection,Taiyuan, China| Shanxi Provincial Key Laboratory for Translational Nuclear Medicine andPrecision Protection, Taiyuan, China

1Radiological and Environmental Medicine
Department of China Institute for Radiation
Protection, China

1

China Institute for Radiation Protection,
Taiyuan, China|2Shanxi Provincial Key Laboratory for Translational Nuclear Medicine andPrecision Protection, Taiyuan, China

1

China Institute for Radiation Protection, Taiyuan, Shanxi, China

1

Department of Radiology and Environment
Medicine, China Institute for Radiation Protection, Taiyuan, Shanxi, China| 2School of
Forensics, Shanxi Medical University, Taiyuan, Shanxi, China

Abstracts - Poster List
No.

Title

Authors#7
Radiological
Emergency
P7-14
Planning
and Preparedness

Evaluation of the Portable Radiation Portal
Monitor’s Usability for
Contamination Screening of Large Groups in a
Radiological Emergency

Young Min Lee1,
Hyun Ha Lee1,
Hyon Sock
Chang2,
Tae Hyung Lim2

P8-01

Development optimized
evaluation technology
for work safety linking
location tracking and insitu measurement indecommissioning facilities

Sang Bum Hong1

P8-02

Component analysis forsolidification after incineration alternative thermal treatment processof dry active wastes

Ji Hyun YU1,
Byung Chae LEE,
Jong Hoa KIM,
Jang Soo SUH

Measurement of Thermal Neutron Fluence
Rates of a Proton Therapy Facility Using a Superconducting Synchrocyclotron - in the Caseof IBA ProteusOne at
Sapporo Kojinkai Memorial Hospital, Japan

Hajime
NAKAMURA1,
Hiroshi MATSUMURA1,
Go YOSHIDA1,
Akihiro TOYODA1,
Eisuke WATANABE1,
Kazuyoshi
MASUMOTO1,Taichi MIURA1,
Hirotaka
NAKAMURA2

P8-03

Topic

#8
Radioactive
Waste andAccident Scenario
Current
P8-04 Radiological Analysis in Radioactive
Waste Treatment Facility
IssuesSan Chae,
Won-Hyuk Jang

P8-05

Development of Radwaste Management
System based on ICT
technology

Jongjin Kim,
Tack-Jin Kim,
Jin-Woo Lee

P8-06

A Study on the Packaging Measures for Radioactive Waste of NPP
Decommissioning

Jun Ki Baik1,
Ji Hwan Yu1,
Byong Sop Lim1,
Jin Won Son1

P8-07

Study on the Treatmentof Radioactive SpentResin by Wet Oxidation

Yahui Xi

P8-08

HiroshiMatsumura1,
Go Yoshida1,
Radioactivation InvesAkihiro Toyoda1,
t i g a t i o n i n S y n c h ro Kazuyoshi
tron-Type Proton TheraMasumoto1,
py Facilities
Hajime Nakamura1,
Taichi Miura1,
Takeji Sakae2,
Naoki Kondo3249

Affiliation
1

Dept. of Nuclear Emergency Preparedness,
Korea Institute of Nuclear Safety, Korea|2SI
Detection, Korea

1
Decommissioning technology development
division, Korea Atomic Energy Research Institute, Korea

1

Sae-An Enertech Corporation, Korea

1

Radiation Science Center, High Energy Accelerator Research Organization, Japan|2Sapporo High Functioning Radiotherapy Center
SAFRA, Sapporo Kojinkai Memorial Hospital,
Japan

1

Korea Atomic Energy Research Institute(KAERI), 34057, Daejeon, Rep. of Korea

1

KAERI (Korea Atomic Energy Research Institute), Korea

1
Central Research Institute, Korea Hydro &Nuclear Power (KHNP), Daejeon, Korea

1

China Institute for Radiation Protection, China

Radiation Science Center, KEK, Japan|2Proton Medical Research Center, University ofTsukuba, Japan|3Medipolis Proton Therapy
and Research Center, Medipolis Medical Research Institute, Japan1

Abstracts - Poster List
No.

Topic

Title

Authors

P8-09

Analysis of Low-Level
Radioactive Concrete
Waste using Gamma
Spectroscopy applying
with Efficiency Transfer

Kyunghun Jung,
Minyoung Jung,
Woojin Jo,
Jung Bo Yoo

P8-10

Preliminary study of theeffect of dominant Co60 in gamma spectrumof radioactive waste

Sooyeon Lim1,
Kahee Jeong1,
Ji-eun Park1,
Wook Sohn1

Dose Assessment forRegulatory Clearance ofOrganic Liquid Radioactive Waste at KAERIKi Baek SHIN1,
Hyun Myoung
JUNG1,
Dong Su KIM1,
Tack Jin KIM1,
Gyu Seong CHO2

P8-12

Degumming test of radioactive filter element

Yongtai Yuan,
Wei Xu,Bowen Zheng,
Peng Wang,
Yu Zhang,Zhijie
Bai,Jiashen Ruan

P8-13

Recommendations for
the Supervision of Mobile Radioactive Waste
Incinerator

Su Ziqiang,
Dong Yuyang,
Chen jiachen,
Lian Bing

In-situ evaluation of neutron flux using activatedreaction of 55Mn(n,γ)56Mn, 23Na(n,γ)24Na
on the component of atandem accelerator

Go YOSHIDA1,
Hiroshi
MATSUMURA1,
Hajime
NAKAMURA1,
Taichi MIURA1,
Akihiro TOYODA1,
Kazuyoshi
MASUMOTO1,
Takayuki
NAKABAYASHI2,
MakotoMATSUDA3

P8-15

Derivation the activity of
129I limit for near surface disposal

Yangjun Zhao1,
Jie Yang1

P8-16

Comparison of Cost about
Equipment for Spent
Nuclear Fuel Management
by Dry Storage Methodin Korea

Hyeok Jae Kim1,
Shin Dong Lee1,
Min Woo Kwak1,
Kwang Pyo Kim1

P8-17

Analysis of Recyclability of Concrete Wastes
Generated from Nuclear
Facilities during Decommissioning

Ji Ung Kim1,
Tae Young Kong2

1
Department of Nuclear Engineering, Chosun
University, Gwangju, Korea| 2Departmentof Nuclear Engineering, Chosun University,
Gwangju, Korea

P8-18

Radiochemical analysis
for difficult-to-measureradionuclides in radioactive samples

Myung Ho Lee1

Young Jae Maeng, KRIST, South Korea|2Mi
Joung Lim, KRIST, South Korea|3Choon Sung
Yoo, KRIST, South Korea|4Byoung Chul Kim,
KRIST, South Korea

P8-19

Establishment of radioactive waste full cycleh i st o r y m a n a g e m e n tsystem using QR-code

Jin-Woo Lee,
Jong-Jin Kim,
Dong-Ju Lee

P8-11

#8
Radioactive
Waste andCurrent
P8-14 Radiological
IssuesAffiliation
1

Korea Atomic Energy Research Institute,
Korea

1

Central Research Institute of KHNP, Korea

1

Korea Atomic Energy Research Institute,
Korea|2Korea Advanced Institute of Scienceand Technology, Korea

1

China Institute for Radiation Protection, China

1

China Institute for Radiation Protection, China

1High Energy Accelerator Research Organization (KEK), Tsukuba-Ibaraki, Japan|2Japan
Radioisotope Association, Tokyo, Japan|3Japan Atomic Energy Agency, Tokai-Ibaraki,
Japan

1

China Institution for Protection, Shanxi Taiyuan, 030006, China

1

Department of Nuclear Engineering, Kyung
Hee University, Korea

1

250

1

KAERI (Korea Atomic Energy Research Intitute), Korea

Abstracts - Poster List
No.

P8-20

Topic

Title

Authors

A Case Study on the Removal and Disposal of
RPV from the La Crosse
Boiling Water Reactor
Decommissioning Experience.

Jihwan Yu,
Hyung-woo Seo,
Gi-lim Kim

Adsorption Kinetics of
Key Nuclides on GeoP8-21 #8
technical Media Form
Radioactive Reprocessing Plant
Waste andCurrent
Study on Rock-Soil DisRadiological persion of Vadose Zone
P8-22 Issues
and Aquifer in Typical
Uranium Mining Area inNorthern Guangdong

P8-23

Mössbauer Investigation of Low Carbon
Steel-Sodium Bentonite
Mixture Aged by γ-Rays
Irradiation

Affiliation
1

Korea Hydro & Nuclear Power (KHNP) Central Research Institute, Republic of Korea

Tian Xie1

1
China Institute For Radiation Protection,
China

Zhao Ming
WANG1,
Song YANG1,
Wen Jie CHEN1,
Chao Chen1,
Ai Ming ZHANG1

1
China Institute for Radiation Protection, China

1

Dong LIANG

251

Department of Radioactive Waste Management, China Institute for Radiation Protection，China

Abstracts - Poster

P1 - 01#1 Radiation Transport and Shielding

Operation test for the muon beam transport system of the μSR
facility at RAON using a beta radiation source

Kihong Pak1, Jae Chang Kim1, Jae Young Jeong1, Yong Hyun Kim1, Yong Kyun
Kim1,*
Department of Nuclear Engineering, Hanyang University/Korea1
Email: ykkim4@hanyang.ac.kr
Keywords: Muon Spin Rotation/Relaxation/Resonance, Muon Beamline, BeamTransport Test, Beta Radiation Source
The first μSR facility in Korea has been constructed at the RAON accelerator
complex in Daejeon. This facility utilizes surface muons (μ+), generated throughinteractions between high-energy proton beams and a graphite target, to study
the material properties of experimental samples by delivering and injecting muons
into them. The μSR facility comprises a muon production system, a beam dump
system, and a beam transport system.
The muon beam transport system consists of various types of electromagnets,
including a pair of normal-conducting solenoids for collecting surface muons,
two dipoles with a deflection angle of 70° for bending the trajectories, nine
quadrupoles, and a Wien filter for rotating the muon spin angle by 45°. The Wien
filter also terminates particles other than surface muons with a selected velocity,
utilizing its orthogonal electric and magnetic fields.
The operation test of the muon beamline was carried out after the installation andalignment of the electromagnets, excluding the Wien filter, were completed.
Since the proton beam is not yet available, and thus the muon beam is also
unavailable, an alternative method using a beta radiation source was chosen. The
beta source, 106Ru/Rh, emits beta particles with a maximum energy of 3.55 MeV.
The target momentum to be delivered by the beamline was set to be 1.69 MeV/c(K.E = 1.25 MeV), due to its maximum intensity in the spectrum.
The first and second focal points were determined by beam optics calculations
applied to surface muons with a momentum of 28.5 MeV/c, and the focal points
were identically applied to the beta particle transport experiment. A double-sided
silicon strip detector (DSSSD) with position resolution was located at the first
focal point, and the experimental setup was configured with a preamplifier, ashaper with CFD, an ADC, a coincidence and gate generator, and a VME bridge.
The strength of the magnetic field for each electromagnet was decreased
according to the ratio of the magnetic rigidity of the beta particle (Bρ_β≈ 0.0056

252

#1 Radiation Transport and Shielding

P1 - 01Tm) to that of the surface muon (Bρ_μ≈ 0.0951 Tm), and the polarities were
switched to the opposite.
Beta particles injected into the first focal point were detected after performing
fine-tuning of the magnetic fields, and the beam size was estimated to beapproximately 16 × 19 mm² with about 4 cps. Subsequently, beta particles weredetected at the second focal point. Although the beta particles were delivered tothe detector, focusing was not achieved. The inability to achieve focusing canbe attributed to: 1) increased fluctuations due to the application of 6% of theactual operation current, 2) inaccuracy of the applied magnetic field caused by
the mismatch between the power supplies used by the manufacturer to measure
the BI curve and those used in the actual facility, and 3) non-standardization of
the utilized power supplies. It is expected that more accurate measurements will
be achievable once these uncertainties are resolved for actual operations using
muons in the future.
Acknowledgement
This work was supported by the Rare Isotope Science Project of the Institute for
Basic Science funded by the Ministry of Science and ICT and the NRF of Korea(2013M7A1A1075764).

253

Abstracts - Poster

P1 - 02#1 Radiation Transport and Shielding

Validation of Nuclear Data Libraries Using Radiation Shielding
Benchmarks in SINBAD

Haelee Hyun1,*, Do Heon Kim1, Ser Gi Hong2
Korea Atomic Energy Research Institute, Daejeon, Korea1, Department of
Nuclear Engineering, Hanyang University, Seoul, Korea 2
Email: hyunhl@kaeri.re.kr
Keywords: SINBAD benchmark, Radiation shielding benchmark, Nuclear datalibrary
The SINBAD (Shielding INtegral Benchmark Archive Database) project, started inearly 1990’s, is a collaborative effort between the OECD/NEA Data Bank
(Organization for Economic Cooperation and Development, Nuclear Energy
Agency Data Bank) and ORNL/RSICC (Oak Ridge National Laboratory, Radiation
Safety Information Computational Center). The main objective of the project is topreserve the information on the performed radiation shielding benchmark
experiments and provide the standardized database for the validation of radiationtransport computer codes and nuclear data libraries. The database is including
101 benchmark experiments and categorized into three parts, reactor shielding(47), fusion neutronics shielding (31), accelerator shielding (23).
To validate and compare the performance of the nuclear data, we used therecently available evaluated nuclear data libraries of ENDF/B-VII.1, ENDF/BVIII.0, JEFF-3.3, JEFF-4T2 and JENDL-5. A total of 26 benchmark problems has
been selected from each category: 7 for reactor shielding, 14 for fusion
neutronics shielding, 5 for accelerator shielding. The benchmark calculations arecarried out using the MCNP6.2 code.
For the validation of each nuclear data library, we analyzed the neutron leakage
spectra from several benchmark problems, including the IPPE-Th in the reactor
shielding category, the IPPE-V and IPPE-Fe in the fusion neutronics shielding
category, and the HIMAC-Fe in the accelerator shielding category. The IPPE
experiments observed transmitting 14 MeV neutrons through a shell made ofthorium (IPPE-Th), vanadium (IPPE-V) and iron (IPPE-Fe). The neutron leakage
spectrum from each shell was measured by TOF (Time-Of-Flight) method. In the
HIMAC experiment, high energy neutrons (up to 800 MeV) were measured as they
penetrated iron shields. The neutron leakage spectrum penetrated through ironshields were measured using the self-TOF detector.

254

#1 Radiation Transport and Shielding

P1 - 02

The benchmark calculation results showed slight differences from theexperimental results in the neutron leakage spectra depending on the nuclear
data libraries used. However, it was confirmed that all of nuclear data libraries
effectively demonstrated the neutron leakage spectra obtained from theexperimental results.

255

Abstracts - Poster

P1 - 03#1 Radiation Transport and Shielding

Study of transformation of HTO in lettuce for atsmosphericrelease of tritiated water

Qiao Cao1,*, Jianguo Li, Shaofei Cao
Institute of Nuclear Environmental Science，China Institute for Radiation
Protection,China1
Email: 2436314992@qq.com
Keywords: tritiated water(HTO)；atsmospheric release；lettuce
Tritified water (HTO) released to the atmosphere mixes with air moisture andexchanges with water pools in plants,soil,and animals.Plants absorb tritiatedwater (HTO) mainly through two ways:one is the absorption of plant leaves fromthe atmosphere,the other is the absorption of plant roots from the soil.In order tostudy the migration and transformation behavior between plants and theenvironment of the release of gaseous tritiated water (HTO),lettuce was selected
as the experimental plant.
Simulation research experiments were carried out in the radiation ecology
laboratory.The tritium concentrations in the cabin air and soil as well as TFWT
and OBT in plants was measured after each experiment.
Conduct HTO pollution experiments simultaneously in 12 small sealed
experimental chambers,and select one of them every 4 hours to conduct
sampling and analysis for the air,plants,and soil in the chamber.During theexperiment,the experimental lettuce was in the mature stage.The results showed
that:(1)For different exposure experimental groups,the concentration of tritiatedwater in the air and soil water varied greatly.The concentration of tritiated waterin the air was one order of magnitude higher than that in the surface soil.The
concentration of tritiated water in the air water ranged from 10072.66 to 23944.57
Bq·kg-1,and the concentration of tritiated water in the soil water ranged from
816.99 to 2077.43 Bq·kg-1.The concentration of tritium in the air and soil
moisture in the exposure chamber is mainly related to factors such as the rate ofevaporation of tritiated water on the electric furnace,the degree of absorption
and transformation of tritiated water by plants in the chamber,the proportion oftritiated water condensed on the exposed chamber wall,and the temperature andhumidity in the experimental chamber.(2)The concentration of TFWT in lettuce
leaves varies from 4758.70 to 18954.23Bq·kg-1.The percentage ratio of TFWT
concentration in rapeseed leaves under different cumulative exposure conditions
to TFWT concentration in equilibrium conditions ranges from 44.37% to256

#1 Radiation Transport and Shielding

P1 - 03196.88%.(3)After the end of the exposure experiment,the lettuce in theexperimental chambers were transferred to indoor conditions for continued
growth.The residue ratio on the first,second,and third days after contamination
is 0.35,0.14,and 0.027.The residual time of OBT in lettuce is longer,with a residue
ratio of 0.27-0.31.By the third day,the experimental results still have nosignificant loss.

257

Abstracts - Poster

P1 - 04#1 Radiation Transport and Shielding

Assessment of Dose Calculation Algorithms by Monte Carlo
Method(MCNP)

Changyeon Yoon1,*
KHNP Central Research Institute, Daejeon/Korea1
Email: changyeon.yoon@khnp.co.kr
Keywords: MCNP, NPP, Dose calculation
To calculate exposure dose for workers, it is necessary to obtain detailed
information of source which causing exposure. However, during the complicatedwork, the work area will change continuously. For this reason, it is difficult toobtain accurate source information, an algorithm based on Electric Power
Research Institute(EPRI) was developed to calculate workers’ exposure dose byusing limited information. In this research, the developed algorithm has beenvalidated by various method.
For the EPRI algorithm, warning statement is shown to the user if error range is
50 to 100%, and a rejected statement is presented to user if error range is morethan 100%. Therefore, it is a reliable result if error range is between 0 to 50%.
The same error condition was used since the algorithm was developed based on
EPRI algorithm.
To verify developed algorithm, Monte Carlo N-Particle Transport Code(MCNP)
which utilized widely was used. The simulation conditions are as follows. One
source was used and consists of Co-60 30%, Co-58 50% and Ag-110m 20%.
There were two measuring points to calculate exposure dose and one shielding
wall was existed. There are 4 points to calculate the radiation dose rate.
In addition, developed algorithm was also verified by experiments. Radiation dose
rates were measured at 430mm, 1330mm, 2230mm, 3130mm and 4030mm
respectively, and measured radiation dose rates compared with calculated
radiation dose rate by using developed algorithm.
Developed algorithm data compared to MCNP were in the range of 0.03~2.28%
relative error. In other words, developed algorithm represent high accuracy not
only simulation but also experiment.

258

#1 Radiation Transport and Shielding

P1 - 05#1 Radiation Transport and Shielding

Analysis of Dose Assessment Methodology by Boundary Voxel of
Irregular Shape Source Based on Point Kernel Method

Min Seong KIM1, Ju Young KIM1, Hyeok Jae KIM1, Kwang Pyo KIM1,*
KyungHee University1
Email: kpkim@khu.ac.kr
Keywords: Radiation Worker, Dose assessment, Point Kernel, Boundary Voxel
Radiation workers in nuclear power plants can be exposed to various shapes ofsources. Point kernel method is a dose assessment methodology that can quickly
and accurately assess the radiation worker exposure dose by splitting sources tovoxels. The voxels are assumed to be point source, and exposure dose is derived
by summing the doses of each voxel. In the case of simple shape sources, it isrelatively easy to assess exposure dose by applying the general point kernel
method. However, in the case of irregular shape sources, the dose assessment
method of boundary voxels should be additionally considered. In actual
workplaces, there are many irregular shape sources, and the appropriate dose
assessment methodology for boundary voxels of sources should be considered.
Therefore, in this study, we analyzed the dose assessment methodology byboundary voxel of irregular shape source based on point kernel method.
In general, the point kernel method performs dose assessment based on a scale
factor. The scale factor is the ratio of the size of the source and the distance
between the source and the exposed person. When the divided voxel satisfies
the criteria for the scale factor specified in advance, the voxels can be assumed
as a point.
In the case of boundary voxels, the centroid and radioactivity of each voxel arecalculated considering the ratio occupied by the source within the voxel. The
methods include 1) random number generation method, and 2) subdivisionvoxels method. The random number generation method identifies random points
in the source of voxels after generating random numbers. Then, the point closest
to the exposed person assumed as the centroid of sphere, and divide the sphere
into 8 divisions as point sources considering scale factor. the average point of
the random points existing in the division sphere is calculated as the centroid ofpoint source, and the ratio of the random points in boundary voxels are calculated
as radioactivity. Finally, the methods are performed repeatedly for considering all
boundary voxels. The subdivision voxel methodology is based on subdividing
boundary voxels. The average points of the voxel in source is selected as the259

Abstracts - Poster

P1 - 05

centroid, and the ratio of subdivision voxels within the boundary voxel iscalculated as the radioactivity.
When applying the point kernel method by irregular shape sources, a moreaccurate dose assessment can be performed by applying the appropriate
boundary voxel dose assessment methodology. It is judged that the results ofthis study can be used when performing exposure dose prediction and evaluation
to establish work plans for workers in actual workplaces.
Acknowledgement
This work was supported through the National Research Foundation of Korea(NRF) using the financial resource granted by the Ministry of Science and ICT(MSIT). (No. RS-2022-00143994).

260

#1 Radiation Transport and Shielding

P1 - 06#1 Radiation Transport and Shielding

Evaluation of neutron effect for multi slit collimator in proton mini
beam radiation therapy
Chanil Jeon1, Sangwoon Jeong1, Youngyih Han2, Sungkoo Cho2,*
Department of Health Sciences and Technology, SAIHST, Sungkyunkwan University, Seoul 06355, Republic of Korea1, Department of Radiation Oncology,
Samsung Medical Center, Sungkyunkwan University School of Medicine, Seoul,
Republic of Korea2
Email: tjdrn39@gmail.com
Keywords: Minibeam radiation therapy, multi slit, neutron, Monte-Carlo
Proton minibeam radiation therapy (pMBRT) is spatial fractionation radiotherapy
technique which has normal tissue sparing effect. However, the technique utilizing multi slit collimators (MSC) made of high density materials generates extra
neutrons that deliver unnecessary irradiation dose in surface area. Therefore, investigation of neutron effect is necessary to apply the technique in clinical fields.
In this study, the effect of neutron was evaluated with according of MSC’s material and thickness using Geant4 10.7 Monte-Carlo simulation toolkit.
To assess neutron effect by MSC, average neutron energy in air box of 300 x 300
x 1 mm3 at target surface was calculated using phase space and depth dose by
neutron was scored in high density polyethylene solid water phantom of 30 x 30
x 20 cm3. In previous research, MSC was designed and optimized to draw effective peak-to-valley dose ratio through parameter adjustment of MSC's block,slit size and thickness for 150 MeV proton wobbling beam with 3 cm spread-out
Bragg Peak using Geant4 toolkit. Under the same conditions, the neutron effectaccording to the material of MSC was evaluated for tungsten, aluminum, iron,and brass, and the effect according to the thickness of MSC was evaluated for40, 60, 80, and 100 mm.
The average energy of neutron for tungsten, brass, iron, and aluminum was 5.0,
10.2, 11.2, and 18.6 MeV, respectively. However, the tungsten MSC delivered thehighest dose in the surface region of the solid water phantom. The difference ofneutron average energy between each thickness of MSC was not significant, andthe MSC of thickness 100 mm delivered lowest dose to the solid water phantom.
As a result, brass was highest efficiency in candidate substance of MSC, andthicker MSC had lower neutron effect to the target.
Based on analysis of the simulation results, the neutron effect of pMBRT using
MSC will be measured in the future study.
Acknowledgement
This work was supported by National Research Foundation of Korea grant funded
by Korean government(MSIT). (No. 2021ME2E8A1048108)
261

Abstracts - Poster

P1 - 07#1 Radiation Transport and Shielding

Research on 3D Radiation Field Calculation Based on CAD
Modeling

Zeyu Long*, Qingliang Feng, Peitao Song, Yuan Zhao, Hua Li
China Institute for Radiation Protection, China1
Email: 1435033781@qq.com
Keywords: Point kernel integration, radiation field, CAD modeling
In nuclear facilities, workers often need to enter radiation areas to perform various
tasks. Obtaining the distribution of the 3D radiation field within these areas inadvance can greatly help optimize radiation protection for workers. Point kernel
integration, Monte Carlo simulation, and discrete ordinates method are commonlyused methods for calculating radiation fields, among which point kernel
integration stands out for its fast computational speed, making it the mostfrequently employed technique for large-scale radiation field calculations.
However, using the point kernel integration method requires the scenario's
calculation model to be modeled in advance. The complexity of scenarios in
actual nuclear facilities makes modeling work challenging. Although CAD
software is flexible and convenient in modeling work, the point kernel integration
method cannot directly use models generated through CAD software. Integrating
CAD software with the point kernel integration method can effectively improvethe efficiency and accuracy of the calculation model construction.
This paper proposes a 3D radiation field calculation method based on CAD
modeling. Firstly, the initial calculation model, which comprises five solid models(cylinders, cones, cuboids, spheres, and round tubes) and four surface models(rectangular surfaces, circular surfaces, spherical surfaces, and cylindricalsurfaces), is modeled using CAD software and exported in the STEP file format.
Then, a custom-developed program module is employed to automatically identify
the model types, extract and compute the corresponding coordinates,
dimensions, and rotational angles from the STEP file based on its data structure,and perform the reconstruction of the calculation model. Lastly, thecorresponding material types or source item data are incorporated into thecalculation model, enabling the 3D radiation field calculations to be carried out
for the radiation area.
After testing the proposed method with designed cases, the reconstructedcalculation model exhibited only slight differences from the original calculation
model, which were attributed to computer floating-point arithmetic errors, and262

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P1 - 07

the deviation of the radiation field calculation results caused by this can beignored, demonstrating the efficacy and reliability of the proposed method.
Acknowledgement
The authors appreciate the fund and supported from the fourth Science Fund forTalented Young Scholars (No. 29) of the China National Nuclear Corporation(CNNC) and the Equipment Advance Research Project (32104040303).

263

Abstracts - Poster

P1 - 08#1 Radiation Transport and Shielding

Radiation Shielding and Activation Analyses for Neutron Activation
Measurement Facility at ARTI, Jeongeup

Jounghwa Lee*, Do Heon Kim, Tae-Yung Song
Korea Atomic Energy Research Institute, Daejeon, Korea1
Email: leejh033@kaeri.re.kr
Keywords: Neutron activation cross section, Radiation shielding, Activation
analysis
Nuclear data is essential for research and development in the field of nuclear
engineering. Nuclear reaction data can contribute to the improvement of nuclear
reaction models as well as the production of medical isotopes and various
radiation applications. In addition, there is a growing demand for research on
fusion materials which need to be further improved. For this purpose, themeasurement of neutron activation cross section using the 30 MeV cyclotron at
Advanced Radiation Technology Institute (ARTI) is planned.
The neutron activation measurement device consists of a target, a graphite
stopper, and a cruciform chamber containing them. The target will be bombarded
with 30 MeV proton beam from the cyclotron to generate neutrons, which will be
used to activate samples such as Au, Co, and Nb to measure the neutron
activation cross sections. Radiation from primary and secondary particles
generated during the experiment was expected to be low enough not to require
an additional shielding design.
In this study, the radiation shielding and activation analyses were performed forthe neutron activation measurement facility at ARTI in Jeongeup. The Li-7 or Be9 are considered as target materials, but the calculations were done only with
Be-9, which yields more neutrons. The irradiation of 30 MeV, 1uA proton beam
was assumed as the operating condition of the cyclotron. MCNP6.1 was used forradiation transport calculations, and FISPACT 2010 was exploited for associateddepletion and activation calculations. The dose conversion factors for neutronsand photons were taken from ICRP-74.
For radiation shielding analysis, dose maps were drawn to investigate the
distribution of radiation in the facility during the measurement of the neutron
activation cross section. In addition, dose rates were calculated at thelabyrinthine entrance and key locations around the facility. The total dose rates
for each location do not exceed the radiation dose limit prescribed in theregulation.

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The activation analysis was performed for equipment, interior wall concrete, andair inside the facility. The types of radioisotopes produced during 1000 hours ofbeam irradiation and their radioactivities were calculated. In addition, the dose
rates for workers near the neutron target system after 1 hour of operation werecomputed.
Acknowledgement
This work was supported by KAERI Institutional Program (Project No. 52441023).

265

Abstracts - Poster

P1 - 09#1 Radiation Transport and Shielding

Monte Carlo optimization design of simulated workplace neutron
fields for typical PWR workplaces

Faguo Chen1,*, Hui Li1, Deyuan Li1
China Institute for Radiation Protection, China1
Email: chenfaguo1@126.com
Keywords: simulated workplace neutron field, Monte Carlo simulation, D-T neutron source
The response of neutron dosemeter is strongly dependent on neutron energy,
spanning from thermal neutron to more than ten MeV for most nuclear facilities.
In practice, neutron dosemeters are usually calibrated in broad spectrum reference neutron radiation fields produced with radionuclide sources, which havedifferent energy distributions with realistic workplaces, resulting in significant
measurement errors. According to ISO 12789, the use of simulated workplace
neutron field having similar energy spectrum with workplace, is an effective alternative method for calibration of neutron dosemeters.
Simulated workplace neutron fields for typical workplaces of pressurized water
reactors (PWR) are planned to be established based on a D-T neutron source
surrounding with modular moderator assembly, which can be installed and adjusted to different targeted neutron spectrum conveniently. This paper describes
the optimization design of moderator assembly and the characterization of corresponding neutron field using Monte Carlo calculations.
As referred in IAEA Technical Reports Series No.403, neutron spectra measured
at PWR pumps room and reactor hall were selected as targeted spectra, withaverage neutron energy of 0.102 MeV, 0.218 MeV and fluence-to-ambient dose
equivalent conversion coefficients h*(10) of 67.71 pSv•cm2, 110.52 pSv•cm2,
respectively. The D-T neutron source was approximated as isotropic emission
with a Gaussian distribution of average energy of 14.1 MeV. Low Z materials suchas water, heavy water, graphite, polythene and high Z materials such as iron,
lead, tungsten and depleted uranium, were used as candidate moderators. The
moderator assembly was designed as a main moderated hemisphere, an adjusting layer, and a reflective shell to regulate fast, intermediate, and thermal neutrons respectively. At the reference point 50 cm away from the assembly along
the axial center, Monte Carlo calculations were used to characterize the neutronenergy spectrum distribution, average energy and conversion coefficients h*(10)under different moderator materials and thickness configurations.

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For PWR pumps room, when moderator assembly consists of 8 cm tungsten, 30
cm iron hemisphere and 30 cm long polythene reflective shell, the simulated
neutron spectrum is close to the measured spectrum, with average energy of
0.139 MeV, conversion coefficient h*(10) of 64.53 pSv•cm2, but an obvious discrepancy near 20 keV. For PWR reactor hall, moderator assembly of 8 cm tungsten, 10 cm iron hemisphere and graphite adjusting layer with a radius of 25 cm
and a thickness of 5cm, can result in a closer conversion coefficient h*(10) of
109.68 pSv•cm2, but a larger average energy of 0.585 MeV.

267

Abstracts - Poster

P1 - 10#1 Radiation Transport and Shielding

Improvement of integrated functions of the dose rate model forincident-free transportation of radioactive materials

Kiwon Jang*, Ara Go
Department of Radiation Protection &Radioactive Waste Safety, Korea Instituteof Nuclear Safety, Korea1
Email: k647jkw@kins.re.kr
Keywords: Transportation, Radiation Risk, Radioactive Waste,
As the demand for transportation of is expected to increase due to decommissioning of nuclear power plants in Korea, the integrated computer code is currently under development to assess the radiation risk associated with thetransport of radioactive wastes including spent fuels for the purpose of regulatory
verification, based on dose models implemented by commonly-used computercodes such as RADTRAN and MARINRAD as references.
RADTRAN is a computer code to evaluate the radiological impact from the transportation of radioactive materials under the both incident free and accident conditions. It was first developed by Sandia National Laboratory (SNL) to support the
U.S. Nuclear Regulatory Commission’s (NRC) environmental impact analysis andRADTRAN6 version is currently being managed by the U.S. NRC. The RADTRAN6
technical manual provides the neutron dose rate model to calculate the integrated
dose to members of the public due to incident free transportation. The integrated
dose to persons surrounding the transport link (off-link) and travelling along thetransport link (on-link) are expressed using Bessel functions in the RADTRAN6
technical manual. In this study, the mathematical improvements in these integrated equations were founded, which are modification of the coefficients andexponential term, and it was verified using benchmark calculation. When comparing the results calculated using MATLAB code with the improved equations inthis study and those calculated using the RADTRAN6, it was found that the
RADTRAN6 results are underestimated.
These results will be utilized when developing the computer code customized tothe environmental characteristics of Korea to evaluate radiological consequences
and risk associated transportation by land or marine of low and intermediate level
radioactive wastes and spent fuels.

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P1 - 11#1 Radiation Transport and Shielding

Effects of Dose to Remaining Buildings from the Nuclear Power
Plant Decommissioning by Shielding Materials and Thickness

Hyung-Woo Seo*, Gang-Woo Ryu
Decommissioning Technology Group/KHNP CRI1
Email: sehywo1727@khnp.co.kr
Keywords: Building reuse, RESRAD-Build, Decommissioning
Dose assessment of buildings that will remain on the site after the decommissioning of the Nuclear Power Plant (NPP) is usually performed using the RESRADBuild code. There are several input parameters in the RESRAD-Build code fordose evaluation, of which shielding factors are ignored for conservative evaluation in buildings. However, it will be necessary to consider the effect of shielding
when wearing protective equipment to protect workers. Therefore, in this study,
the effect of shielding materials and thickness provided by the RESRAD-Build
code was evaluated applying the potential radionuclides of contaminations on
the buildings.
The selection of scenarios, input parameters, deterministic or probabilistic evaluation methods, etc. is necessary to evaluate the dose of exposed persons affected by reuse buildings using the RESRAD-Build code. Scenarios for building
occupancy and renovation can usually be applied. Among the input parameters,
the shielding factor was changed so as to evaluate its effect in addition to thecommon values.
In general, for radionuclides in which gamma emission is dominant, the effect ofexternal exposure is greater than that of internal exposure. Therefore, the thickness effect of the shielding is greater for gamma-emitting radionuclides. In the
results of the resident farmer scenario, the decay effect of Co-60 was the largest,and then the effect of Sr-90 was also identified. In terms of the effect of the twoshielding materials, it was confirmed that the shielding effect was also significant
because the lead had a high density.
In decommissioning nuclear facilities, safety evaluation of remaining buildings
whose licenses have been terminated is an essential issue. Usually, the Derived
Concentration Guideline Level (DCGL) should be calculated to verify that thecontamination inside the building meets the site reuse criteria (i.e., 0.1 mSV/yr).

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P1 - 11To this end, the use of RESRAD1Build as a computer code considers the shielding effect negligent for conservative evaluation. However, it may need to checkhow much the shielding effect is, and it was confirmed by increasing the thickness of two shielding materials for three representative radionuclides. In thisstudy, the basic results of shielding were derived in advance. Therefore, moredetailed analysis of various assumptions about shielding will be considered infuture studies.

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P1 - 12#1 Radiation Transport and Shielding

Research and Verification of Radiation Shielding for a Nondestructive Testing Device of a Coal Mine Belt

Jianyu Fu1, Xiangming Xue1, Jingming Zhan1,*
China Research Institute of Radiation Protection1
Email: 936388580@.com
Keywords: Nondestructive testing equipment， Dose equivalentrate，Radiological protection， Shielding effect
X-rays are a double-edged sword. The human body often suffers from different
degrees of pathological changes due to receiving excessive doses of radiation.
The excessive dose of ionizing radiation hinders and damages the function ofhuman cells, seriously leads to cell death, and causes damage to human cells
and tissues. With the increase of radiation dose, the damage becomes more andmore serious.
X-ray was first discovered by the German physicist Roentgen more than 80 years
ago. Its discovery has touched the world. With the development of the times, Xray inspection has played an important role in non-destructive testing of steel
wire rope core belts. On the one hand, it can clearly identify the belt damage, on
the other hand, it can damage the body of the belt operator. Therefore, sincesuch devices are used in the coal mine, it is the responsibility to inform thepersonnel exposed to radiation of the hazards of exposure, and provide themwith corresponding personal protective equipment, so as to reduce the dose ratelevel of radiation workers.
Investigate the surrounding dose rate level during the operation of the belt NDT
device in a coal mine in Shanxi Province, and estimate the annual effective dose
received by the belt operator. Leverage X γ The dose rate instrument (FH40G10+FHZ672E-10) measures the dose equivalent rate around the device,compares the measured data with the requirements of the dose control value in
the Radiation Protection Standard for Industrial Flaw Detection (GBZ117-2022),
uses the design shielding to carry out the radiation protection of the NDT device,
and verifies the radiation shielding effect by measuring the dose rate level aroundthe device after the shielding protection is installed. The results show that theshield body of the belt nondestructive testing device is reasonably designed,
which can effectively shield the X-ray generated after the device is turned on,
and make the exposure dose of workers exposed to radiation reach a controllable
level.

271

Abstracts - Poster

P2 - 01#2 Radiation Dosimetry

Analysis of usefulness of obesity risk factors using ROC curve formale workers
Baek Yun-Mi*, Kang Seo-gon, Kim Jeong-in, Sung Sook-hee
Korea Hydro & Nuclear Power Co., Ltd1
Email: ympaek2000@khnp.co.kr
Keywords: Male worker, Obesity, ROC Curve, Translocation chromosome aberration<<<Purpose<<<DNA damage includes not only changes at the nucleotide level, but also damage
and chromosomal abnormalities at the chromosomal level. It can be caused byoxidative stress and exposure to endogenous and exogenous factors (chemicals,
drugs, toxins, ultraviolet rays, ionizing radiation) and can occur naturally inhealthy people.
Obesity induces DNA damage due to oxidative stress and inflammation, and theaccumulation of DNA damage alters gene expression and can interfere with cell
metabolism. Other inflammatory diseases increase oxidative stress and are associated with chronic systemic inflammation and chromosomal damage.
This study aims to identify the factors related to the presence or absence ofobesity in male workers and to analyze the usefulness of the markers that canbest predict obesity using the ROC curve.
<<<Methods<<<
This study was conducted after prior approval from the Clinical Research Ethics
Committee of Hanil Hospital (HGH –2014-HDM-002). Obesity among male workers who voluntarily consented to participate in the research, aged 30 to 59 years
old, who worked in the industry from January 2015 to December 2020. Treatment
guidelines (2020) A total of 194 patients were enrolled: 79 normal(18.5≤BMI≤22.9) and 115 obese (BMI≥25).
For the data survey, a self-administrated questionnaire, health examination, peripheral blood collection for chromosome evaluation, culturing for 48 hours, andstable chromosomal aberration frequency after harvest were evaluated by counting 1,000 samples.
Statistical processing was performed using Windows SPSS 26.0 (SPSS Inc., Chicago, IL, USA), differences between groups were analyzed by one-way ANOVA,
associations between variables were analyzed by correlation analysis, and optimal obesity markers were determined. ROC (Receiver Operating Characteristic)
curve was used for this, and the statistical significance level was set when the Pvalue was less than 0.05.
272

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P2 - 01<<<Results<<<
There was no statistically significant difference in age between the normal group(42.1 years) and the obese group (44.0 years) (P=0.213)
Among the clinical indicators, statistically significant variables were BMI (kg/m2)
normal group (mean 21.6), obese group (mean 27.4) (P=0.000), Abdominal circumference (cm) normal group (mean 80.8), obese group (mean 93.5).
(P=0.000),
GPT normal group (mean 31.5), obese group (mean 25.8) (P=0.043), r-GTP normal group (mean 46.8), obese group (mean 32.9) (P=0.014), Insulin normalgroup (mean 7.4), obese group ( mean 5.4) (P=0.000), HbA1c normal group(mean 5.3), obese group (mean 5.1) (P=0.000), Fasting plasma glucose (mg/dL)
normal group (mean 103.5), obese group (mean 97.6) (P = 0.000), translocation(yield per 1,000 cells), normal group (mean 3.5), and obese group (mean 4.5)
(P=0.042).
As for the correlation between obesity indicators, waist circumference had thehighest correlation (0.850), and TR (0.227) was second.
, and the lowest correlation was GOT (-0.135)
As a result of using the ROC (Receiver Operating Characteristic) curve to determine the optimal obesity marker, the first was the waist circumference, with an
AUC of 0.922 (0.884 –0.961) and a cut-point of 84.50 or higher, which was themost useful marker. Translocation (yield per 1,000 cells) was the second largest
area at 0.60 (0.484 - 0.659), and the cut-point was over 1.50.
<<<Conclusion<<<Waist circumference and stable chromosome variation were found to be major
factors in the occurrence of obesity in male workers. Since the number of subjects is limited to generalize, further research is needed in the future.
Acknowledgement
This study was conducted as part of the KHNP (A23LF01) and the Ministry ofIndustry's research project (20224B10300090).
Reference
1. Armstrong, A.; Bohannan, B.; Mburu, S.; Alarcon, I.; Kasparek, T.; Toumi, J.;
Frade, S.; Barrio, S.F.; Augustin, M. Impact of Psoriatic Disease on Quality of
Life: Interim Results of a Global Survey. Dermatol. Ther. 2022, 12, 1055–1064 2.
Dobrică, E.C.; Cozma, M.A.; Găman, M.A.; Voiculescu, V.M.; Găman, A.M. The
Involvement of Oxidative Stress in Psoriasis: ASystematic Review. Antioxidants
2022, 11, 282 3. K. Makki, P. Frogel, I. Wolowczuk, adipose tissue of obesityrelated inflammation and insulin resistance: cells, cytokines and chemokines, andISRN Inflamm. 2013, 139239. 4. Niazi, Y.; Thomsen, H.; Smolkova, B.; Vodickova,
L.; Vodenkova, S.; Kroupa, M.; Vymetalkova, V.; Kazimirova, A.; Barancokova,M.;
Volkovova, K.; et al. DNA repair gene polymorphisms and chromosomal aberrations in healthy, nonsmoking population.DNA Repair 2021, 101, 103079.
273

Abstracts - Poster

P2 - 02#2 Radiation Dosimetry

Development of n-γ mixed fields based on a D-D neutron sourcefor validation of material-filtered RPLGD for BNCT

Zixu Xu1,*, Kazuma Aoki1, Shingo Tamaki1, Sachie Kusaka1, Isao Murata1
Graduate School of Engineering, Osaka University1
Email: Xu.Zixu@qr.eng.osaka-u.ac.jp
Keywords: Mixed field, radiation dosimetry, energy spectrum, MCNPIn the treatment field of accelerator-based Boron Neutron Capture Therapy(BNCT), gamma-rays are easily produced during neutron irradiation. In the BNCT
project of Osaka University, material-filtered Radio-Photoluminescence Glass
Dosimeters (RPLGD) are developed to measure neutron and gamma-ray doses
separately and simultaneously. In this study, to validate the material-filtered
RPLGD for BNCT, various types of n-γ mixed fields were designed based on a
D-D neutron source.
An irradiation assembly was designed to generate n-γ mixed fields, including a
D-D neutron source (isotropic, 2.5 MeV) and a multi-material moderator (cylinder:r = 20 cm). Six materials were considered to design the moderator: Al (2.70g/cm3), Fe (7.86 g/cm3), Pb (11.34 g/cm3), PE (0.93 g/cm3), CF2 (2.20 g/cm3)
and AlF3 (2.88 g/cm3). Because n-γ mixed fields were generated after gammarays production by incident neutron moderation, the gamma-ray to neutron dose
ratio, ηd, was able to be controlled between 0% and 200%. The calculations were
performed by Monte Carlo N-Particle Code 5 (MCNP5). When D-D source neutrons penetrate a neutron-generating target, gamma-rays are produced by capture and inelastic scattering reactions, which is evaluated as a contaminant contribution in the n-γ mixed field. In this study, the shielding ability of each material
to absorb the gamma-rays emitted from the target was examined. 19 cm Fe and12 cm Pb were discovered to absorb the gamma-rays emitted from the target
with an absorption ratio of exceeding 99.5%. In contrast, 0-60 cm Al, 0-70 cm
AlF3, and 0-70 cm CF2 exhibited a lower shielding ability, with an absorption
ratio of up to 98.7%, 95.3%, and 97.7%, respectively. PE was used to produce
more gamma-ray dose and adjust ηd between 0% and 200%.
The results were summarized as follows: (i) Fast neutron dominant field: PE (024 cm) + Fe (19 cm), with ηd from 15.5% to 188.1%, proportion of fast neutronfrom 88.0% to 99.8%, and gamma-ray energy up to 10.1 MeV; (ii) Epithermal
neutron dominant field: PE (5 cm) + Fe (1-5 cm) + AlF3 (44 cm) + Pb (1-4 cm),with ηd from 25.3% to 204.4%, proportion of epithermal neutron from 59.7% to274

#2 Radiation Dosimetry

P2 - 0269.2%, and gamma-ray energy up to 10.0 MeV; (iii) Thermal neutron dominantfield: PE (29 cm) + Pb (12 cm) + PE (1-7 cm), with ηd from 32.9% to 192.9%,
proportion of thermal neutron from 62.1% to 69.1%, and gamma-ray energy up
to 6.1 MeV.
It was confirmed that the present study provides valuable reference n-γ mixed
fields for developing material-filtered RPLGD for BNCT.
Acknowledgement
This work was supported by China Scholarship Council (No.202106240024) and
JSPS KEKENHI (No. JP19H02648).

275

Abstracts - Poster

P2 - 03#2 Radiation Dosimetry

Dosimetric comparison of different techniques of intraoperative
radiotherapy for spinal metastasis based on MC simulation

Junyi Liu1, Miao Qi1, Ning Gao1, Yikai Wu1, Xi Pei3, X.George Xu2,*
School of Nuclear Science and Technology, University of Science andTechnology of China/China1, Department of Radiation Oncology, First AffiliatedHospital of University of Science and Technology of China/China2, School of
Nuclear Science and Technology, University of Science and Technology ofChina/China2, Anhui Wisdom Technology Company Limited/China3
Email: xgxu@ustc.edu.cn
Keywords: Monte Carlo, Radiation Dosimetry,Intensity modulation,
Intraoperative radiotherapy(IORT)
Purpose: Intraoperative radiotherapy (IORT) for spinal metastases can beclinically delivered using a mobile 50-kV X-ray generator in the surgical room,
aided with the vertebral arch puncture involving a metal needle applicator.
However, complex anatomical structures make it difficult for the applicator toreach the target center. The healthy lungs and aorta often times are inevitably
irradiated by isotropic sources that are designed to deliver a uniform radiationdose to the entire target area. We propose an intensity modulation dose methodthat is commonly done with the “external beam radiotherapy”. This study carries
out Monte Carlo simulations of intraoperative radiotherapy for spinal metastases
involving different intraoperative radiotherapy delivery methods including
traditional isotropic dose methods and the proposed intensity modulation
methods.
Methods: Three different radiation sources (50-kV Intrabeam, 50-kV micro X-ray
tube and 40-kV micro X-ray tube) are considered for intraoperative treatment. A
theoretical applicator, having shapes such as the angular shield structures
reported in the recent literature, are modeled using Monte Carlo simulation
methods. Isotropic irradiation and intensity modulation irradiation based onshielding for all three sources are simulated for comparison. The geometricposition of these radiation source is maintained in all plans, and the target
minimal dose of 8 Gy is required for all plan. Computed tomography (CT) patient
data are processed and imported into TOPAS, a Monte Carlo user code basedon the generic code GEANT4, for dose calculations. AI based auto-segmentation
software Deepviewer was used for contouring of organs at risk. Dose-volume
histograms are derived for all cases for further dose evaluation.

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Results: All treatment plans are found to meet the basic clinical treatment planrequirements. The CTV dose parameters are equivalent for all isotropic plans.
Among intensity modulation plans, the CTV dose parameters of Intrabeam and
50-kV micro X tube are equivalent, and the CTV D95% of 40kV micro X tube isslightly worse than that of Intrambeam plan. For all isotropic plans, comparedwith the Intrabeam plan, the doses of proximal lung and aorta of 40-kV /50-kV
micro X tube plan are reduced by varying degrees, and the improvement of the
40-kV micro X tube plan is greater. All intensity modulation plans show significant
improvement in V5Gy and D2cc of proximal lung compared with isotropic plans.
Conclusions: In this study, Monte Carlo simulations are performed to compare
traditional and proposed dose plans. Dosimetric parameter analyses of different
intraoperative radiotherapy methods show that the proposed intensity modulation
methods are advantageous in protecting the organs at risk while ensuring target
coverage.

277

Abstracts - Poster

P2 - 04#2 Radiation Dosimetry

Development of Computed Tomography-dedicated Mesh-Type
Pediatric Phantom Library: Arms-removed and Arms-raised
PostureSuhyeon Kim1, Bangho SHIN1, Chansoo CHOI2, Hyeonil KIM1, Haegin HAN1,
Sungho MOON1, Gahee SON1, Chan Hyeong KIM1,*
Hanyang University, Seoul, Korea1, J Crayton Pruitt Family Department of
Biomedical Engineering, University of Florida, Gainesville, FL, USA2
Email: chkim@hanyang.ac.kr
Keywords: mesh type phantom, phantom library, CT dose, posture deformation
The increased use of recent diagnostic computed tomography (CT) examinations
for children and adolescents has raised the need to optimize the radiation dose
while maintaining image quality, considering the high radiosensitivity and longer
life-expectancy of the pediatric patients. For this purpose, the pediatric computational human phantoms are commonly used to estimate the radiation doses.
However, the pediatric phantoms are generally developed in the arms-down position, which is not consistent with the typical arms-raised CT examination posture. This inconsistency could lead to inaccurate dose estimation. To address
this issue, in the present study, CT-dedicated (i.e., arms-removed and armsraised posture) mesh-type pediatric phantom library was developed based on
the body-size-dependent pediatric phantom library (i.e., arms-down posture)
covering a wide range of body sizes and ages. The arms-removed phantom library was developed simply by deleting the arms of the arms-down phantom
library. The arms-raised phantom library was developed by updating and applying
the in-house automated posture deformation program, which was used to develop the CT-dedicated adult phantom library, to the arms-down pediatric phantom library. To see the dosimetric impact of the CT-dedicated pediatric phantom
library, some arms-raised and arms-removed phantoms were implemented into
the Geant4 Monte Carlo radiation transport code to calculate red bone marrow(RBM) and lung doses per unit milliampere-seconds for a chest helical CT examination, which were then compared with the values calculated from the armsdown phantoms. For the RBM, the arms-removed phantoms showed the highest
values, and the arms-raised phantoms showed the lowest values. These results
are due to the fact that the RBM of the arms-removed phantoms is directly exposed to the radiation (i.e., without shielded from the skin or cortical bone) at

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the arms-removed cross section while the RBM included in the arms for thearms-raised phantoms were not exposed to the radiation. For the lung, the armsremoved phantoms still showed the highest values, whereas the arms-down
phantoms showed the lowest values. These results are due to the shielding effectof the arms. The CT-dedicated pediatric phantom library is expected to be usedin various fields such as CT dose estimation, CT optimization for radiation doseand image quality, and radiation therapy.

279

Abstracts - Poster

P2 - 05#2 Radiation Dosimetry

Image Reconstruction with Limited-view-angle Projection Datafor BNCT-SPECT―Establishment of the Mock-up System andValidation of the Response Function―

Fan Lu, Kouki Okuda, Haruka Inamoto, Shingo Tamaki, Sachie Kusaka, Isao
Murata*Department of Engineering, Osaka University/ Japan1
Email: murata@see.eng.osaka-u.ac.jp
Keywords: BNCT-SPECT, image reconstruction, limited-view-angle projection,
mock-up system, response function
Backgrouns and Purpose:
Boron Neutron Capture Therapy (BNCT) is expected to be a new radiation therapy
for cancer, and researches on various issues associated with it are still inprogress. One of the unsolved issues is the construction of a system to observe
the effect of the treatment (local boron dose) in real-time during the treatment.
In particular, there are high expectations for a method to measure the gamma
rays emitted promptly via boron-neutron capture reaction from outside of the
body and to reconstruct the images in a SPECT-like manner (BNCT-SPECT).
94% of 7Li produced by 10B (n, α)7Li reaction is in an excited state (7Li*) andemits a 478keV prompt gamma-ray by the transitions to the ground state. The
number of 478keV gamma-ray, which is proportional to that of 10B (n, α)7Li
reactions, directly indicates the therapeutic effect. However, several difficult
problems left should be solved to develop the BNCT-SPECT system. One of theimportant issues is the projection angle limitation of the measurement device. Ina typical SPECT system, in which the projection angle range of the detector groupis up to about 360 degrees at most. Nevertheless, in BNCT-SPECT, we must set
up the patient's body on the surface of the neutron exit wall, and the angle
available for the radiation measurement is limited to less than around 180 degrees.
Therefore, Fourier image reconstruction methods that require projections from atleast 180 degrees directions cannot basically be applied to BNCT-SPECT.
Methods: Based on the above, we aim to develop a new successive
approximation image reconstruction method based on Bayesian Estimation, valid
even under the severe conditions specific to BNCT-SPECT, and finallydemonstrate the BNCT-SPECT. In this study, we perform mock-up experiments
and MCNP simulations to obtain limited-view-angle projection data and the280

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response function with which we can reconstruct images successfully.
Results: We constructed a mock-up system to investigate experimental
specifications for the BNCT-SPECT system, and to validate the response
function of the measurement system of the BNCT-SPECT evaluated by MCNP5.
However, the mock-up system was not able to reproduce the actual accuracy ofa real BNCT-SPECT system because the source intensity in the mock-up
experiment was too weak compared to the real one. So, we had to establish an
experimental system with the same statistical accuracy and signal-to-noise ratio
as the real BNCT-SPECT system. For this purpose, the mock-up system was
modified by changing the position of gamma-ray sources to reproduce the actual
accuracy of the designed BNCT-SPECT.
Conclusion: We have successfully established an experimental mock-up systemthat could reproduce the actual accuracy, and the response function was verified
with the mock-up system. We can finally develop a reliable image reconstruction
system for BNCT-SPECT.

281

Abstracts - Poster

P2 - 06#2 Radiation Dosimetry

Establishing Local Exposure Scenarios for Local Skin Dose
Assessment System

Seung Beom Goh1, Han Cheol Yang2, Chan Joong Kim1, Jae Chang Kim1,
Kihong Pak1, Yong Hyun Kim2, Yong Kyun Kim1,*
Department of Nuclear Enginnering, Hanyang University/Korea1, Department of
Nuclear Enginnering, Hanyang University/Korea2, Raymetrics, Co./Korea2
Email: ykkim4@hanyang.ac.kr
Keywords: Overexposure on the Hand Skin, Radiological Accident Scenario,
Local Skin Dose Assessment
When radiological accidents happen, one of the main overexposures to organs
is the skin on the hands. After radiological accidents, current Retrospective
Dosimetry (RD) is performed, but current RD techniques such as TL/OSL, EPR,
and biological dosimetry are limited in accurately assessing local radiation
exposure on the skin, particularly on the hands.
To resolve these limitations, research is needed to develop a system forassessing local skin dose on the hands. Since artificially creating radiologicalaccidents for research raises ethical issues, an analyze past accidents which
local skin overexposure occurred for establish radiological accident scenarios fordose assessment.
In this study, a radiological accident scenario with local exposure on the skinwere established based on an IAEA's ‘The Radiological Accident’ publications.
We selected five incidents from the reports published by the IAEA between 2002
and 2018, which included local exposure to the hands, and analyzed each
incident.
The victims of these accidents who were exposed by radioactive sources usedin medical or industrial fields. The activity of the sources in these incidents
ranged from 185 GBq to 1295 TBq. Symptoms of Acute Radiation Syndrome(ARS), such as skin erythema, blisters, fibrosis, inflammation, and necrosis, wereobserved in the hand skin of the victims. The most commonly used method for
RD was biological dosimetry, with EPR dosimetry used in three of the fiveincidents. The absorbed doses received by the overexposure victims in their handskin were approximately 1 to 40 Gy, with the palms and fingers being the mostaffected local areas.
In future work, an ultra-thin scintillator printed by 3D printer and a 3D scanningbased anthropomorphic phantom for local skin dose assessment system will be282

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P2 - 06

developed. Especially, ultra-thin scintillators can be made to the thickness of theskin basal layer (50 μm) which can meet the personal dose equivalent, Hp(0.07)
criteria recommended in ICRP publication 103. Moreover, we will conduct a
reproducible experiment of radiation exposure doses at the time of a radiation
accident using the scenario selected in this study and the system.
Acknowledgement
This work was supported by the Nuclear Safety Research Program through the
Korea Foundation Of Nuclear Safety(KoFONS) using the financial resource
granted by the Nuclear Safety and Security Commission(NSSC) of the Republicof Korea. (No. RS-2023-00242425)
Reference
IAEA, 'The radiological accident in Chilca', (2018), ISBN 978-92-0-101817-5
IAEA, 'The radiological accident in Nueva Aldea', (2009), ISBN 978-92-0103009-2 IAEA, 'The radiological accident in Lia, Georgia', (2014), ISBN 97892-0-103614-8 IAEA, 'The radiological accident in Samut Prakarn', (2002), ISBN
92-0-110902-4 IAEA, 'The radiological accident in Gilan', (2002), ISBN 92-0110502-9

283

Abstracts - Poster

P2 - 07#2 Radiation Dosimetry

Numerical simulation for calculating counting efficiencies
considering the biodistribution of I-131 in hand-held radiation
instrument

MinSeok Park1,*, Jaeryong Yoo1
Department of Emergency Medical Research, Korea Institute of Radiologicaland Medical Sciences/Korea1
Email: chulipak@kirams.re.kr
Keywords: Virtual calibration, Internal contamination, Monte Carlo simulation,
Counting efficiency
After a nuclear accident, radiological release into environment can cause theinternal contamination of the general public. Among them, radioiodine mainly
contributes to the internal dose and should be monitored to mitigate theassociated health risks. Population monitoring is necessary to screen the large
number of potentially contaminated individuals in response to a radiologicalemergency. In the initial phase of an accident, field triage screening should be
conducted using hand-held radiation instruments to assess the internal
contamination of affected populations. However, these instruments are notproperly calibrated for individual monitoring, causing errors in measurement
results. To address this issue, virtual calibration was applied to calibrate theradiation instrument. Through virtual calibration, counting efficiencies can becalculated, considering anatomical differences among individuals andbiodistribution of intake scenarios. The present study aims to calculate the timedependent counting efficiencies of hand-held radiation instruments using virtual
calibration based on Monte Carlo simulation with mesh-type computational
human phantoms.
A NaI(Tl) based survey meter for thyroid measurement was modeled using aMonte Carlo code, and computational modeling was validated based oncomparison between experiment and simulation. Based on age-specificbiokinetic models, the biodistributions of I-131 depending on intake scenarios
were calculated and applied to age- and gender-specific computational humanphantoms for simulating contaminated individuals with different ages andgenders. To maintain the reproducibility of measurement geometry, the distancesand positions between the detectors and computational human phantoms in thesimulation were determined to be the same as the practical measurement284

#2 Radiation Dosimetry

P2 - 07condition.
The results reveal that counting efficiencies could be significantly influenced byanatomical characteristics of individuals and biodistribution of I-131. Counting
efficiencies for pediatrics were larger than those for adults. From the perspectiveof the biodistribution of I-131, counting efficiencies in the early phase after its
incorporation are dramatically changed over time due to its rapid biologicalmovement. After two weeks or more, counting efficiencies become constantbecause the retained iodine is distributed in body tissue and thyroid.
This study performed virtual calibration to calculate counting efficiencies, andthis method can minimize the influence that could occur due to the limitations ofconventional calibration. The time-dependent counting efficiencies calculated inthis study could be practically used in internal dose assessment and can help
provide net counts rates corresponding to the committed effective dose forradiological triage in the field.
Acknowledgement
This study was supported by a grant of the Nuclear Safety Research Program
through the Korea Foundation Of Nuclear Safety (KoFONS) using the financial
resource granted by the Nuclear Safety and Security Commission (NSSC) of
Republic of Korea (No. 2104038).

285

Abstracts - Poster

P2 - 08#2 Radiation DosimetryMesh-Type Phantom Posture Deformer: A Real-Time GUI for
Personalized Phantom Generation

Haegin Han1, Sungho Moon1, Gahee Son1, Bangho Shin1, Hyeonil Kim1, Suhyeon Kim1, Chan Hyeong Kim1,*
Department of Nuclear Engineering, Hanyang University1
Email: chkim@hanyang.ac.kr
Keywords: Mesh-type Phantom, Monte-Carlo, Compuational Phanotm
The International Commission on Radiological Protection (ICRP) recently released
a mesh-type reference computational phantom (MRCP) for adult male and female, offering refined definitions of small or thin organs and high deformability.
Taking these advantages, a phantom library for various body shapes and five
posture-deformed phantoms have been developed to investigate dosimetric impact of individual characteristics. However, these studies have limitations in thatthe body shape and posture were considered as independent factors and only
five representative postures were investigated. In addition, the previous methodology requires a few days to create a new posture-deformed phantom, being
inappropriate for generating the phantoms for numerous combinations of body
shapes and postures. This study presents the mesh-type phantom posture deformer (MPPD), a user-friendly GUI program that offers easy and rapid posturedeformation of the any mesh-type phantoms, including those of phantom library.
For an input original phantom, the MPPD automatically performs the preprocessing computations, including skeleton structure generation and the Bounded
Biharmonic Weights (BBW) calculation, which takes ~2 minute. Once the preprocessing is done, users can modify the phantom posture simply by manipulating the widgets, which triggers the posture deformation by Dual Quaternion Skinning (DQS) algorithm. This calculation was implemented with multithreading capability to enable real-time visualization of the posture-deformed phantom on
the GUI. For the demonstration of the MPPD, an adult male MRCP was deformed
into a kneeling posture and its effective dose was calculated and compared withthat the existing kneeling MRCP. The phantom was naturally deformed even in
the sophisticated spine and neck areas, and the total volume was increased by~4%, which is due to the inherent bulging effect of the DQS algorithm. When theeffective doses for parallel photon beam in AP-direction were compared, thedose values for the two phantoms showed good agreement within 10% of thediscrepancies in the energy range of 0.001-10,000 MeV. The MPPD developedin the present study is expected to be useful in various research fields which
requires personalized phantoms.
286

#2 Radiation Dosimetry

P2 - 09#2 Radiation Dosimetry

Deep Learning for Stable Chromosome Analysis in Bio-Dosimetry

LEE In Kyung*, HAN Jung Hun
RHI, Korea Hydro & Nuclear Power Co., LTD1
Email: krakow1102@khnp.co.kr
Keywords: Deep learning, Stable chromosome analysis, Bio-dosimetry
Counting the number of chromosomes and detecting abnormal chromosomes
could be crucial steps in diagnosing diseases and inferring the scale of radiation
exposure. Moreover, the analysis of translocations in stable chromosomes isused as the most useful method due to the relative stability of their frequency
with post-irradiation time. to assess the frequency of stable chromosomal
aberrations (i.e., translocations), whole-chromosome probe painting was
performed for chromosomes 1,2, and 4, which represent 22.06% of the humangenome.
As we adopt algorithms for deep learning for stable chromosome analysis, image
processing techniques using features are found in various forms, either
overlapping or touching and separating them is a challenging task. So, we
attempted to simplify the separation of overlapping chromosomes and expand
the range of detection of abnormal chromosomes.
In this study, two algorithms, counting and detection systems, were developed
based on YOLOv5 and trained with improved YOLOv5 for stable chromosome
analysis. And short of the dataset in abnormal images, poor performance has
resulted. Therefore, data augmentation was conducted to prevent overfitting andimprove the model's performance as an effective method. We tried numerous
augmentation methods, including the spatial-level transform, which changes theimage itself, pixel-level transform, which changes the units of the pixel, thecutout method, which cuts out a part of images, and the mosaic method, whicharranges four images into a 2x2 array. Moreover, stable chromosome images
have RGB color channels that have different intensities, and noise levels could
be detectable harder than unstable chromosome images. So, we pre-treatedstable chromosome images like red-gray color using UNet and ResNet bae. UNet
is a convolutional neural network(CNN) architecture that was developed forbiomedical image segmentation and is very effective for our tasks with the ResNet
backbone. The object detection algorithm generates bounding boxes with theSoft-NMS method, which lowers the confidence score of the overlappedb(bounding)boxed. It has been proposed to leave the number of normal and287

Abstracts - Poster

P2 - 09abnormal chromosomes unbalanced, attempts to resolve imbalances, such as
replacement of the loss function or data upsampling, should be performed inparallel.
This study was performed as a pilot within a short period of time, but we have
tried many models and seen the possibility. And it is expected that this study will
reveal that the performance of chromosome analysis is improved in terms of theevaluation indicators.

288

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P2 - 10#2 Radiation Dosimetry

Deep Learning for Unstable Chromosome Analysis in
Bio-Dosimetry

LEE In Kyung*, KIM Jeong In, BAEK Yoon Mi, KANG Seung Young
RHI, Korea Hydro & Nuclear Power Co., LTD1
Email: krakow1102@khnp.co.kr
Keywords: Deep learning, Unstable chromosome, Bio-dosimetry
The traditional approach for chromosomal analysis is time-consuming and laborintensive, and since all work is done manually, mistakes in the expert operator's
experiment and analysis are unavoidable. Therefore, several automated algorithms were developed to eliminate errors even in the most critical analyses andmake operators; jobs easier. However, in practice, the accuracy was very low,
and most of the chromosomal images could not be analyzed for abnormal chromosomes.
Examples of unstable chromosome aberration include dicentric, tricentric, ring,
acentric, fragment, etc. Although radiation induces many types of chromosomalchanges, dicentric chromosomes are considered the most sensitive and specificfor assessing radiation dose.
In this study, we developed the chromosome counting system and abnormal
chromosome detection system in unstable chromosomes based on deep learning.
Both systems adopted the object detection method and were designed based on
YOLOv5. And to improve the extraction of chromosome features, spatial andchannel attention mechanisms were enhanced to YOLOv5. Also, both systems
were trained with improved YOLOv5. In the case of radiation-exposed chromosomes, the frequency of overlapping chromosomes is high, and the overlapping
objects should be separated. Therefore, the counting system divided overlapping
objects via the soft non-maximum suppression (Soft-NMS) method. In the caseof detecting abnormal chromosomes, the improved YOLOv5 that added various
attention techniques to extract specific features as abnormal chromosomes are
structurally different from normal chromosomes effectively detected abnormal
chromosomes.
Counting and detecting systems generally detect chromosomes according totheir purpose, however, they have some limitations. In the counting system, whennumerous chromosomes overlap, when small chromosomes overlap invisibly, or
when chromosomes are too close, chromosome detection is not properly performed. In addition, in the detection system, small chromosomes were mistakenly

289

Abstracts - Poster

P2 - 10judged as fragments and folded chromosomes as dicentrics.
To overcome these problems, the diversity of data must be ensured in the future.
And for the counting system, images with adjacent chromosomes and images
with many overlapping chromosomes must be learned. For the detection system,
both normal and abnormal chromosomes must be learned to prevent misrecognition.

290

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P2 - 11#2 Radiation Dosimetry

Enhancing Radiologist Pose Tracking in Interventional Radiology:
A YOLOv7 and Multi-Camera Approach for Virtual Dose Monitoring System

Sungho Moon1, Haegin Han1, Gahee Son1, Bangho Shin1, Hyeonil Kim1, Suhyeon Kim1, Soon Young Song2, Chan Hyeong Kim1,*
Hanyang University1, Hanyang University Hospital2
Email: chkim@hanyang.ac.kr
Keywords: Interventional Radiology, Pose Tracking, Monte Carlo
Recently, the research team developed an initial model of the Virtual Dose Monitoring System for Interventional Radiology (VIR), a system designed to computethe radiation dose of the interventional radiologist in real-time. One of the coretechnologies of VIR, the body tracking module, currently obtains the radiologist
poses through the Azure Kinect (Microsoft Corp., USA) depth camera and a dedicated pose tracking program provided by Microsoft Corporation. This approach
offers the advantage of acquiring 3-D data with a single camera; however, it
presents a significant challenge in pose tracking accuracy that may occur by
obstructions and/or several staff. To overcome this limitation, in the present study,
a novel radiologist pose tracking methodology was developed, employing multiple color cameras and the YOLOv7 algorithm, which demonstrates considerable
stability to occlusion. The YOLOv7 algorithm, based on deep learning, detects
humans in color images and tracks the positions of their body joints. Since thetracked joint positions are 2-D form within the image, a minimum of two images
with joint position data is necessary to reconstruct the coordinates in 3-D space.
In this study, the images from the Nutz UHD 2160 webcams (PSI Korea, South
Korea) were used and the 3-D joint position coordinates were calculated by Direct Linear Transformation (DLT) algorithm. Subsequently, the rotation data foreach bone, required for the posture deformation of the phantom, were determined based on the 3-D joint position coordinates. The developed methodology
was then compared to the previous methodology by tracking the pose in thesame scenes and accordingly deforming the posture of the mesh-type referencecomputational phantom (MRCP). To compare dose differences between bothphantoms, the Geant4 Monte Carlo toolkit was used to calculate the effective
dose for general interventional procedural conditions. It was observed that themethodology developed in the present study reliably tracked the radiologist poses
despite the occlusion, while the pose tracked via Azure Kinect occasionally
showed distortions. For the frame in which the pose was well-captured in bothmethodologies, the dose discrepancies were within 5%.
291

Abstracts - Poster

P2 - 12#2 Radiation Dosimetry

Development of GUI-based Monte Carlo Simulation Code forExternal Exposure (McSEE): Implementation of Advanced
FunctionsHyeonil KIM1, Bangho SHIN1, Chansoo CHOI2, Suhyeon KIM1, Haegin HAN1,
Sungho MOON1, Gahee SON1, Chan Hyeong KIM1,*
Department of Nuclear Engineering, Hanyang University, Seoul, Korea1,
J Crayton Pruitt Family Department of Biomedical Engineering, University of
Florida, Gainesville, FL, USA2
Email: chkim@hanyang.ac.kr
Keywords: ICRP, Mesh-type phantom, Monte Carlo simulation, Dose calculation,
Graphical user interface (GUI)The International Commission on Radiological Protection (ICRP) formulated Task
Group (TG) 112 with the aim of providing the reference datasets and methodologies for emergency exposure situations. Acknowledging the advantages of the
ICRP mesh-type reference computational phantoms (MRCPs), e.g., phantom
deformability and the definition of radiosensitive target layers, TG 112 decided
to provide reference dose coefficients (DCs) for external exposures using the
MRCPs. Considering the practical difficulties in providing DCs for every possible
exposure scenario, however, TG 112 determined to provide DCs for representative exposure scenarios. For the other cases, TG 112 decided to provide a reference Monte Carlo code, thereby enabling the radiation workers or regulators toeasily and accurately calculate DCs using the ICRP MRCPs by themselves. In ourprevious study, a GUI-based Monte Carlo Simulation Code for External Exposure(McSEE) was developed by implementing the four basic panels: phantom, geometry, source, and output. To expand the coverage of the code for external
exposure situations and enhance the convenience for users, in the present study,
the McSEE was improved by implementing advanced functions in the phantom,
geometry, and source panels. The new version includes an extended list of phantoms, such as the ICRP pediatric MRCPs, the body-size-dependent pediatricphantoms, and the user-defined arbitrary posture phantoms created using theMesh-type Phantom Posture Deformer (MPPD) program. To account for scenarios involving several workers or the public, multiple phantoms can now be implemented in the McSEE. Moreover, the phantoms, along with the wearables, canbe translated and rotated intuitively by using the mouse control. Next, users can292

#2 Radiation Dosimetry

P2 - 12now define wall thickness for shielding objects to create various geometries andset transparency levels to view the inner side of these objects. Lastly, the McSEE
allows the implementation of the phase-space file, in the form of the International
Atomic Energy Agency (IAEA)- and user-defined files, to generate complex radiation sources. Once fully developed, the McSEE will be used by TG 112 to
calculate and provide the reference DCs for representative emergency exposure
situations and will be distributed through an upcoming ICRP Publication to calculate DCs for various external exposures by the general radiation workers or
regulators.

293

Abstracts - Poster

P2 - 13#2 Radiation Dosimetry

Evaluating Fetal Radiation Exposure: Development and Impact ofPregnant-female Mesh-type Reference Computational Phantoms
for Accurate Dosimetry

Bangho Shin1, Chansoo Choi2, Suhyeon Kim1, Hyeonil Kim1, Haegin Han1,
Sungho Moon1, Gahee Son1, Yeon Soo Yeom3, Beom Sun Chung4, Chan
Hyeong Kim1,*
Department of Nuclear Engineering, Hanyang University, Seoul, Korea1, J
Crayton Pruitt Family Department of Biomedical Engineering, University of
Florida, Gainesville, FL, USA2, Department of Radiation Convergence
Engineering, Yonsei University, Wonju, Korea3, Department of Anatomy, Yonsei
University Wonju College of Medicine, Wonju, Korea4
Email: chkim@hanyang.ac.kr
Keywords: International Commission on Radiological Protection, pregnantfemale, mesh-type reference computational phantom, specific absorbed
fraction
The International Commission on Radiological Protection (ICRP) emphasized theimportance of accurate assessment of radiation doses to fetus due to its
substantial lifetime cancer risk associated with fetal exposure. For this purpose,through ICRP 2007 recommendation, the ICRP decided to develop and provide
reference pregnant-female phantom series, consisting of maternal and fetal
phantoms. Meanwhile, acknowledging the advantages of the tetrahedral-meshphantoms, the ICRP developed the adult and pediatric mesh-type referencecomputational phantoms (MRCPs) by converting the existing voxel-typereference phantoms. In 2020, the ICRP decided to provide the reference
pregnant-female phantom series, i.e., fetal ages at 8, 10, 15, 20, 25, 30, 35, and38 weeks, only in mesh format and provide the specific absorbed fractions (SAFs)
for fetus as a target. As part of this project, the present study will introduce thedevelopment procedures of the pregnant-female MRCPs. The maternal
phantoms were constructed based on the adult female MRCP, with adjustments
made to account for pregnant-female anatomy (e.g., change in rib cage
dimensions). The abdominal organs of the pregnant-female phantom series of
University of Florida (UF) were then converted into high-quality mesh format andinstalled in the maternal phantoms. The fetal phantoms were constructed basedon the newborn MRCPs and the fetal phantom series from UF and Ferdowsi

294

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P2 - 13

University of Mashhad (FUM). The posture of the newborn MRCPs was deformed
into the posture of fetus and the UF and FUM fetal phantoms were converted intomesh format. Then, the fetal anatomy (e.g., anthropometric parameters) wasreflected in the fetal phantoms. Finally, the fetal phantoms were implemented into
the uterus of the maternal phantoms. To see the dosimetric impact of thepregnant-female MRCPs, the SAFs for maternal liver as a source and entire fetus
as a target for fetal ages at 10-, 25-, and 38-week. The result showed that thelarger SAFs were observed with fetal growth. The maximum difference was by afactor of six at the lowest energy (0.05 MeV) between the 10- and 38-weekphantoms. This result is attributed to the reduced distance between the maternalliver and the fetus due to the fetal growth. In the near future, the pregnant-female
MRCPs will be used to calculate SAFs for maternal and fetal organs as sourcesand entire fetus or individual fetal organs as targets. Then, the pregnant-female
MRCPs and the SAF values will be distributed through the upcoming ICRP
publication.

295

Abstracts - Poster

P2 - 14#2 Radiation Dosimetry

Development of a prototype TL/OSL reader for on-site application
in a large-scale radiological accident

Hyoungtaek Kim1,*, Chang-Young Park2, Sang In Kim3, Min Chae Kim1, Jungil
Lee1, Jang Lyul Kim1
Radiation Safety Management Division, Korea Atomic Energy Research Institute,Daejeon, Korea1, Dept. of Physics, Gyeongsnag National University, Jinju,
Korea2, ARIM Science, Daejeon, Korea3
Email: kht84@kaeri.re.kr
Keywords: Retrospective dosimetry, Thermoluminescence, Optically stimulated
luminescence, portable reader
In the event of a large-scale radiological accident, the ability to quickly andaccurately measure radiation doses is essential for effective emergency response
and medical treatment. The ICRU-94 report suggests reliable retrospective
dosimetry techniques applicable at the initial stage of large-scale radiation
emergency. Personal belongings, such as surface mounted devices and display
glasses in a mobile phone, and smart chip cards in a credit and subscriber
identity module cards, are fortuitous materials in dose estimation of individuals
using thermoluminescence (TL) and optically stimulated luminescence (OSL)
methods. These techniques provide high measurement throughput with reliable
results, making them suitable for triage situations where mass casualties are
expected. To date, however, dose assessment is accompanied by sample
collection and transport from on-site to a laboratory, which is expected to be abottleneck for mass measurements.
In this study, we developed a prototype TL/OSL reader for fortuitous dosimetersthat can be applicable on-site. The reader was designed to be portable, easy touse, and capable of reading multiple types of materials. The device consists ofa PMT (photomultiplier tube), a sample heating block, a heating control module,a sample moving stage, an N2 gas control module, a DAQ (data acquisition andreception unit), and a PC (control computer). The sample heating plate is madeof kanthal, and the temperature of the heating plate is controlled by the control
unit and a thermo-coupler. The heating module was designed to heat a sample
from room temperature to 600 °C with variable heating rates. The sample moving
stage was chain driven to allow easy sample mounting and smooth movement to
the PMT. When the reader is in operation, the PMT was blocked from external
light and light from the sample heating module.

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We conducted laboratory tests to evaluate the performance of the reader,including dose recovery of fortuitous dosimeters. The results demonstrate thatthe prototype TL/OSL reader have a potential to be a valuable tool for rapid andaccurate radiation dose measurement in the event of a large-scale radiologicalaccident, and may help to improve emergency response and medical treatment
in such situations.
Acknowledgement
The study was carried out under the National Research Foundation of Korea (NRF)
grant funded by the Korean government (Ministry of Science and ICT) (RS-202200144350, RS-2022-00144210) and KAERI (Project number 521410-23).

297

Abstracts - Poster

P2 - 15#2 Radiation Dosimetry

Evaluation of shielding effect of personal dosimeter according toworking postures

Min Chae Kim1,*, Hyoungtaek Kim1, Sora Kim1, Yoomi Choi1, Jiyoon Kim1,
ByungIl Min1, Kyungsuk Suh1, Jungil Lee1, Haegin Han2, Chan Hyeong Kim2
Korea Atomic Energy Research Institute (KAERI) / Republic of Korea1,
Department of Nuclear Engineering, Hanyang University / Republic of Korea2
Email: minchaekim@kaeri.re.kr
Keywords: Monte carlo simulation, accident dosimetry, personal dosimeter,
shielding effect, working postures
In general, a dosimeter is, in principle, worn on the chest of a radiation worker,
and it is considered to be representative of an exposed dose of an individual.
However, in the event of a radiation accident, workers can be exposed in a varietyof postures due to their movements and emergency works. Therefore, the recorded doses on the chest can be significantly affected by a part of the body suchas torso, arms, and legs.
In a previous study, dose conversion factors were calculated using a mobile
phone as a fortuitous dosimeter for three postures; standing, kneeling and squatting. When the mobile phone was placed on the chest, the difference of doseconversion coefficient between the standing and squatting posture was up to 44%
due to shielding by arms under lateral exposure with Cs-137. Since the postures
used in the previous studies were only three types (including standing) and doses
were calculated using fortuitous dosimeters, it is necessary to calculate exposure
doses reflecting actual working postures with a personal dosimeter.
In the present study, the evaluated doses of personal dosimeter at the chest
position were compared according to various postures to confirm the shielding
effect by the body. The applied computational anthropomorphic phantoms weremesh-type reference computational phantoms (MRCPs) provided by international
commission on radiological protection (ICRP). The MRCPs were modified to havedifferent postures, such as walking, sitting, bending, kneeling, and squatting,considering workplace behaviors. Three radiation sources were used: Ir-192,
Cs-137, and Co-60. Exposure geometries were antero-posterior (AP), posterior-anterior (PA), left lateral (LLAT), right lateral (RLAT), rotational (ROT), andisotropic (ISO). Thermoluminescence dosimeter (TLD) was designed as a personal detector, which was attached to the left chest of the phantoms.

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Acknowledgement
The study was carried out under the National research Foundation of Korea (NRF)
grant funded by the Korea government (Ministry of Science and ICT) (RS-202200144350, RS-2022-00144210).

299

Abstracts - Poster

P2 - 16#2 Radiation Dosimetry

Dose Evaluation of High-energy Photons Leaking from Kindai
Reactor

Shintaro KIMURA1,*, Hirokuni YAMANISHI2
Grad.Sch.Kindai Univ1, AERI,Kindai Univ2
Email: 2233340473s@kindai.ac.jp
Keywords: Dose evaluation,UTR-KINKI,PHITS,BGO scintillation detector
The Kindai University Reactor (UTR-KINKI) is a zero power reactor with amaximum thermal power of 1 W designed for education and research purposes.
The dimensions of the biological shielding tank, which is the external shape ofthe reactor, are 4 m in diameter and 2 m in height. The size of the core region
are 1.2 mL×1.5 mW×1.2mH. The upper part of the core is covered by a concreteof 0.50 m in thickness, and the sides of the core are surrounded by a biologicalshielding tank filled with light water and sand of 1.0 m in average thickness.
Neutrons generated in the core interact with the reactor structure and shielding
materials to produce secondary gamma rays, which penetrate the thin shieldingon the roof of the reactor building and scatter into the air, affecting the dose at
the site boundary as the skyshine component. The power output is 1 W, so thedose rate is low, but it is necessary to evaluate the dose inside the facility andat the site boundary.
In this study, the dose was evaluated by two methods: simulation calculation with
the PHITS (Particle and Heavy Ion Transport code System) and actual
measurement by means of a radiation detector. We used BGO scintillation
detectors (crystal size: 3Φ × 3inch). Due to its large atomic number, the atomic
cross section is large and the detection efficiency is high for high-energy gamma
rays. Measurements were taken for 300 seconds on the top of the reactor (TOR)
at 0.01 W reactor power, 600 seconds on the catwalk (CW) at 1 W reactor power,and 1,800 seconds on the roof of the reactor building (ROB). The size of thereactor room is 20 mL×10 mW×10 mH. The CW is located 6 m horizontally fromthe TOR and 2 m high above the TOR.
Although natural background radiation is below 3 MeV, high-energy photons up
to about 10 MeV were measured. Comparing of the measurement at CW with at
TOR, the counting rate was about 1/100. However, the low energy side showeda higher count rate. This may be due to the fact that TOR directly measures thesecondary gamma rays generated, whereas CW also detects the scatteredcomponent. Comparison of the measured results with previous studies and PHITS

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calculations confirms H-1(n,γ) at 2 MeV, Si-nat(n,γ) at 3 MeV, Ca-40(n,γ) at 6
MeV and Fe-56(n,γ) at 7.5 MeV were identified from the nuclide analysis ofenergy peaks. Both were secondary gamma rays generated by the interaction ofH, Si, Ca, and Fe in reinforced concrete. In the future, dose evaluation by G(E)function will be performed. In addition, we plan to clarify the dose distribution
inside and outside the building using transport calculations with PHITS.

301

Abstracts - Poster

P2 - 17#2 Radiation Dosimetry

Comparing the Performance of an Automatic System with Manual
Scoring in the Dicentric Chromosome Assay

Soo Kyung Jeong1, Su Jung Oh1, Yeong-Rok Kang1, Hyo Jin Kiim1, Chang
Geun Lee1, Song-Hyun Kim2, Seungsoo Jang3, Min Ho Jeong4, Wol Soon Jo1,*
Dongnam Institute of Radiological & Medical Sciences1, KEPCO International
Nuclear Graduate School (KINGS)2, Division of Advanced Nuclear Engineering,
POSTECH3, Dong-A University College of Medicine4
Email: sailorjo@dirams.re.kr
Keywords: Biodosimetry, Dicentric chromosome assay, Automatic scoring
The dicentric chromosome assay is the “gold standard” method in biodosimetry
to estimate the radiation exposure of accident victims to facilitate medical treatment. However, the time-consuming process and the need for trained reviewers
limit the large-scale application. A recently developed automated system that
identifies monocentric and dicentric chromosomes have been evaluated for thedicentric chromosome assay.
In previous work, a deep learning-based automatic dose estimation system(DLADES) was developed to calculate the dose curve using the conventional
statistical method. However, its efficacy compared to manual scoring by humanscorers in the dicentric chromosome assay has yet to be fully explored. Therefore,
this study aimed to compare the performance of the DLADES with manual scoringin the dicentric chromosome assay.
Blood was collected from two donors and irradiated in six dose ranges from 0 to
4 Gy at 0.8 Gy/min with cobalt-60 gamma rays. The DLADES was then used to
identify the dicentric chromosome frequency from the irradiated samples, theresults were compared with those obtained by manual scoring by human scorers.
The results showed that the automatic system produced different results fromhuman scorers, particularly at higher doses. In particular, human scorers' frequency of dicentric chromosomes was about 1.8 times higher at 4Gy. This discrepancy can be attributed to the impact of dose on the accuracy of dicentricchromosome recognition in the automatic system.
As the dose increases, complex chromosomal aberrations also increase, affecting the automatic system's recognition accuracy.
These findings have important implications for automatic systems' use in the dicentric chromosome assay. While the DLADES has the potential to streamline theassay and reduce dependence on trained reviewers, its accuracy at higher doses302

#2 Radiation Dosimetry

P2 - 17needs to be improved to produce more reliable results. Future studies should
explore optimizing the automatic system's performance at higher doses to enhance its efficacy in the dicentric chromosome assay.
In conclusion, the study provides valuable insights into the performance of adeep learning-based automatic dose estimation system in the dicentric chromosome assay compared to manual scoring by human scorers. The findings highlight the importance of dose on the accuracy of the automatic system's recognition algorithm and the need for further research to optimize its performance at
higher doses.

303

Abstracts - Poster

P2 - 18#2 Radiation Dosimetry

Neutron Dose Inside a 50 MeV Cyclotron Vault

Ahn Yunjin1, Kim Gi-sub1, Kim Taehwan1, Kim Sangrok1,*
Radiation Safety section,Korea Institute of Radiological and Medical
Sciences/Korea1
Email: kim@kirams.re.kr
Keywords: CR-39, LR-115, neutron, neutron measurement, high-energy
particle accelerator
A high-energy particle accelerator creates neutron radiation, making it difficult toestimate the dose level. A CR-39 plastic track detector is generally used forneutron dose measurements. The neutron dose measurement limit of the CR-39
detector is known to be 60 mSv. However, neutrons generated in high-energy
particle accelerator facilities, which have recently attracted attention, exceed themeasurement limits of the CR-39 detector. Therefore, an attempt was made tomeasure neutrons using an LR-115 at the Korea Institute of Radiological andMedical Sciences, where a 50-MeV cyclotron is in operation.
For neutron irradiation by the 50-MeV cyclotron, a proton beam was made tocollide with a 0.5-cm-thick beryllium target. Then, a 1-mm aluminum window
through the beryllium target stopped the proton beam. The proton beam, reactingwith the beryllium target, was then converted into a neutron. At that time, theinitial acceleration energies of the irradiated proton beam were 20 and 30 MeV.
The experiment was conducted under a condition of 1–10 µA. The LR-115 was
mounted on a concrete wall 5 m from the beryllium target. After beam irradiation,
the number of tracks per unit area (cm⁻²) of the LR-115 was compared with theneutron dose calculated with a general-purpose Monte Carlo particle transport
code, MCNP6.2.
As a result of 3000 s of proton irradiation under a condition of 20 MeV and 10µA, the number of tracks per unit area on the LR-115 was 314 cm⁻². Comparedwith MCNP calculation, this result was equivalent to 4610 mSv, and a track-todose conversion factor of 14.7 mSv·cm² was derived. The track-to-dose
conversion factors with 300- and 3000-s irradiation of 30 MeV·10 µA were 5.65
and 6.36 mSv·cm², respectively.
This study confirmed that the LR-115 could measure neutrons under high-dose
conditions in high-energy accelerator facilities exceeding the CR-39
measurement range. Therefore, it is concluded that the LR-115 can be used forneutron measurement in high-energy particle accelerator facilities.

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Acknowledgement
This study was supported by a grant of the Korea Institute of Radiological andMedical Sciences(KIRAMS), funded by Ministry of Science and ICT(MSIT),
Republic of Korea. (No. 50422-2023). The authors express their sincere thanks
to the staff of the MC-50 Cyclotron Laboratory (KIRAMS), for the excellent
operation and their support during the experiment.

305

Abstracts - Poster

P2 - 19#2 Radiation Dosimetry

Improvement of the Performance of a Whole Body Counter Using
an Artificial Neural Network

Moonhyung Cho1,*, Hyunjin Kim1
Central Research Institute, Korea Hydro & Nuclear Power Co., Ltd./Korea1
Email: chomh9525@khnp.co.kr
Keywords: Whole body counter, Artificial neural network, Performance testing
In Korea NPPs, whole-body counters (WBCs) are used to measure the radioisotopes (RIs) inhaled by workers. FastScan, a stand-type WBC consisting of two
NaI(Tl) detectors, is mostly used to quantify the amount of inhaled RIs, since it
has a relatively higher efficiency than AccuScan, a bed-type WBC consisting of
one or two HPGe detectors. In 2022, Central Research Institute of Korea Hydro& Nuclear Power (KHNP–CRI) has hosted performance testing for all of the NPPs
in Korea in accordance with the ANSI/HPS N13.30-2011. The performance ofeach NPP is evaluated based on the root mean square error (RMSE): the squareroot of the sum of the square of the relative bias (Br) and the square of therelative precision (SB). The criteria of the performance testing is RMSE ≤ 0.25.
In the case of the lung category, Two RIs among Mn-54, Co-57 or Ce-144, Co58, Co-60 shall be quantified in the presence of interfering RIs (Cs-134 and Cs137). The RMSE values of the NPPs that participated in the performance testing
ranged from 0.05 to 0.4.
In this study, we developed an artificial neural network (ANN) to enhance theperformance of the FastScan and compared the RMSE values evaluated by the
ANN to those obtained through the peak analysis (PA) method of the FastScan.
The architecture of the ANN, resulting in two hidden layers, was determined by
the Bayesian optimization.
To generate the training and validation data sets, we synthesized the normalized
gamma-ray spectra of Mn-54, Co-57, Co-58, and Co-60 because creating datasets by measuring actual spectra for the four RIs would be impractical. We used
50,000 spectra for training and 5,000 spectra for validation. A test set that ismeasured five times over 1 minute for the four RIs is used to evaluate the performance of the ANN.
To evaluate the performance of the ANN, we used a test set composed of fivespectra, each measured for one minute. The RMSE values evaluated by the ANN
ranged from 0.036 to 0.1, which were on average three times lower than thoseof the PA method. The results of our study indicate that the ANN outperformed
the PA method, suggesting that it could be a promising approach to improve theperformance of the FastScan WBC.
306

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P2 - 20#2 Radiation Dosimetry

Uncertainty Evaluation of calibration of a Beta-ray Irradiation
System using the Monte-Carlo Method

Moonhyung Cho*, Hyunjin Kim
Central Research Institute, Korea Hydro & Nuclear Power Co., Ltd./Korea1
Email: chomh9525@khnp.co.kr
Keywords: Beta-ray irradiation system, Uncertainty, Monte Carlo method
Since 2018, Central Research Institute of Korea Hydro & Nuclear Power (KHNP–
CRI) has been operating a beta-ray irradiation system to irradiate thermo-luminescence dosimeters (TLDs) that are used in domestic nuclear power plants(NPPs). The beta-ray irradiation system consists of an irradiator with two Sr90/Y-90 (74 MBq, 1.85 GBq), beta-ray emitting radioisotope (RI), and a rail system. In accordance with ISO-6980-1 “Reference beta-particle radiation-Part 1:
Methods of production”, a beam flattening filter consisting of three concentricdisks mode of polyethylene terephthalate (PET) was installed at a distance of 10
cm from the surface of the RIs. This filter makes it possible to achieve a uniform
dose equivalent rate at a calibration distance of 30 cm from the surface of theRIs.
In this study, we used a Monte-Carlo method to evaluate the uncertainty of calibration of the beta-ray irradiation system
The mathematical model for calibration is defined as the product of five quantities
as follows: the calibration factor of the ion chamber, the current measured by theelectrometer, the accuracy of the electrometer, the correction factor of environmental variation, and the location reproducibility of the ion chamber. The uncertainty of each quantity for the 90Sr/90Y RI with an activity of 74 MBq in the irradiator were calculated to be 1.75 %, 0.335 %, 0.069 %, 0.187 %, and 1.927 %,respectively. The uncertainty in the absorbed dose rate generated by the activityof 1.85 GBq in the irradiator were calculated to be similar to those obtained forthe 74 Mq RI.
Using the Monte Carlo method, the combined standard uncertainty for absorbed
dose rate at a depth of 0.07 mm in ICRU tissue was calculated to be 2.63 % The
coverage factor (k) corresponding to the 95 percent confidence interval was calculated to be 1.9. This value is slightly lower the coverage factor (k = 1.95) for a
Gaussian distribution. This difference seems to result from the uncertainty(1.93 %) of the location reproducibility of the ion chamber, which has a rectangular distribution. In this study, the expanded uncertainty of the beta-ray irradiation system using Monte Carlo method was calculated to be 5.0 %.
307

Abstracts - Poster

P2 - 21#2 Radiation Dosimetry

Research on mRNA as a Biomarker of Mixed Neutron and γ -ray
Irradiation

Yuan Yayi 1, Chai Dongjing1, Dong Juancong1, Jiao Cheng1, Dang Xuhong 1,*
Department of Radiation Medicine and Environmental Medicine, China Institutefor Radiation Protection, China1
Email: dangxuhong005@163.com
Keywords: 252Cf, γ-ray, combined irradiation, neutron, transcriptome

Purpose: Neutron is a kind of high LET ray. Compared with low-LET rays, theradiation damage caused by neutrons is more complicated and difficult to repair,which will cause more serious health effects. In actual scenarios, neutron radiation such as clinical radiotherapy, nuclear power plant leakage, space flight,
high-altitude flight, and atomic bomb explosion is often accompanied by gamma
rays. Because the research on the mixing effect of neutrons and gamma rays is
restricted by irradiation conditions and other factors, the mixing effect of neutrons and gamma rays is still unclear. The study of biological effects is the key
to solve a series of radiation protection problems such as radiation health impact
assessment. Therefore, it is necessary to study the mixing effect of neutron andgamma rays.
Methods: We studied the effects of neutron, neutron+γ and γ-ray combined irradiation on the transcription spectrum in human peripheral blood of 3 healthy
adult men. Samples were irradiated with 1.42 Gy 2.5-MeV neutron irradiation,
0.71 Gy neutron+0.71 Gy 137Cs γ irradiation, and 1.42 Gy 137Cs γ irradiation.Transcriptome sequencing of peripheral blood irradiated by different rays,
screening co-differentially expressed genes and verifying them. The dose-response relationship of AHH-1 cells irradiated by 252Cf was studied.
Results: Transcriptome sequencing identified 56 co-differentially expressed
genes and enriched 26 KEGG pathways. There are 97, 45 and 30 differentiallyexpressed genes in neutron, neutron+γ and γ-ray respectively, and 21, 3 and 8
KEGG pathways with significant differences are enriched. Fluorescence quantitative polymerase chain reaction (qPCR) verified co-differential expression of
AEN, BAX, DDB2, FDXR, and MDM2. Additionally, irradiation of AHH-1 human
lymphocytes with a 252Cf neutron source at 0, 0.14, 0.35, and 0.71 Gy, fluorescence qPCR revealed a dose–response relationship for BAX, DDB2, and FDXR atdose ranges of 0–0.71 Gy, with R2 of 0.803, 0.999, and 0.999, respectively.

308

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P2 - 21

Conclusion: Neutron can induce more differentially expressed genes and enrichmore pathways. Neutron+γ irradiation may combine the damage of high LET and
low LET, the genes activated by neutron+γ irradiation are almost the combinationof genes activated by neutron radiation and γ radiation. BAX, DDB2 and FDXR
are differentially expressed after irradiation by D-D neutron source and 252Cf
neutron source, so they are expected to be molecular targets of neutron damage.

309

Abstracts - Poster

P2 - 22#2 Radiation Dosimetry

Establishment and application of the biokinetic model ofplutonium liver cells

Shuang-Shuang CUI1, Juan-Cong DONG1, Jiao CHENG1, Hong-Yan LIU1,
You-Chen LI1,*
Institute of Radiology and Environmental Medicine, China Institute for Radiation
Protection/China1
Email: liyc69@163.com
Keywords: plutonium, interal exposure dose estimate, biokinetics, liver cells
Plutonium is one of the main occupational hazards of nuclear fuel reprocessing
workers, and also the focus of radiation protection. Liver is the main target organof plutonium. In this study, the transfer process of plutonium in liver cells wasstudied by establishing the method of measuring trace plutonium in cells basedon ICP-MS and the biokinetic model of plutonium liver cells, which provided data
support for evaluating the internal irradiation dose of plutonium in liver target
organs. And Using the established biokinetic model, explore the effect of DTPA
on the transfer of plutonium in liver cells.
ICP-MS can be used to measure trace plutonium, so we choose it to detect trace
nuclides in cells. After culturing normal human liver cell line (L-02) in vitro, we
exposed them to plutonium (242Pu, 0.097Bq/106 cells), collected and counted
the infected liver cells, washed the cells twice to remove the nuclides adheringto the cell surface. The measurement error of ICP-MS is less than 5%, and theexperimental method is repeatable. Based on the established experimental
method for the detection of trace plutonium in liver cells and the growth curve ofliver cells, a liver cells biokinetic model of plutonium was established. The current
model of liver cell biokinetics is to conduct radionuclide exposure for 24h on the
4th day after liver cell inoculation. Measure the uptake of plutonium by liver cells
at 2, 4, 8 and 24 h before the cells were poisoned with+/- DTPA(0.5mM) respectively. The poisoned cells were cultured under +/-DTPA for 2h, 4h, 8h, 24h
and 48h respectively.
Effect of DTPA on plutonium uptake process: The measure results show that theuptake of plutonium by liver cells increased with time and the liver cells in thegroup without DTPA ingested more plutonium than those in the DTPA group. No
DTPA group increased from initial 5.992 ± 0.265pg/106 cells increased to 27.522± 1.431pg/106 cells at 24h, and the DTPA group increased from 3.492 ±0.243pg/106 cells to 6.345 ± 0.557pg/106 cells. Effect of DTPA on plutonium

310

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P2 - 22

release process: The measure results show that the group without DTPA decreased from 89.750 ± 2.662pg/106 cells to 31.530 ± 2.189pg/106 cells, andthe DTPA group decreased from 89.750 ± 2.662pg/106 cells to 21.917 ±
1.862pg/106 cells. The ratio of plutonium content between the DTPA free groupand the DTPA group was 1.256 ± 0.177. The use of DTPA will reduce the uptakeof plutonium by liver cells and increase the excretion of plutonium from liver cells.
In this study, we have established an experimental method for the stable measurement of trace plutonium in cells and the liver cells biological kinetics model ofplutonium, and verified the influence of DTPA on the biological kinetics of plutonium by using the established experimental method and model.

311

Abstracts - Poster

P3 - 01#3 Radiation Detection and Sensor Technology

Development of a Methodology to Convert the Energy Spectra toSpectra for Channel

Ju Hyung Kim1, Yong Hyun Kim2, Ki Hong Park1, Seung Beom Goh1,
Yong Kyun Kim1,*
Department of Nuclear Engineering, Hanyang University, Seoul, South Korea 1,
RayMetrics, Seoul, South Korea2
Email: ykkim4@hanyang.ac.kr
Keywords: data preprocessing, channel, energy
Nuclear species identification using energy spectra of channels in artificial intelligence requires prior learning of target nuclear species. However, acquiring
many kinds of nuclear species spectrum data is challenging, and Monte Carlosimulation is often used to generate spectra. To use MCNPX simulation-obtainedenergy spectra as training data, it is necessary to convert them to spectra forchannels by a calibration function. At that time, the channel values converted by
the converted function are not integers. Thus, the calculated channel values are
not in the format of experimental data, making them unsuitable for application to
training data. If the calculated values are rounded to the nearest integer, thiscause the problem of creating duplicate channels and omitting channels, which
is not appropriate for representing channels of spectra.
To address this problem, in this study, we propose a novel method to transform
MCNPX simulation spectra data into integer channel data. The method involves
defining a sequence of arbitrary channel values, adding new rows with the average count value for rows with non-matching channel values, and reordering therows based on channel index. The process can be repeated across all rows. As
a result, the data preprocessing code successfully transformed the MCNPX simulation spectrum data into count by integer channel without value overlappingand missing. And a comparison between the preprocessed and measured spectra showed consistent peak channel positions, confirming the method's appropriateness.
In conclusion, this study proposes a data preprocessing method to address theissues that arise when using MCNPX simulation data as training data for artificial
intelligence. The proposed method successfully transformed the MCNPX simulation data into integer count data for each channel without damaging the spectrum's shape. This method can be used to prepare training data for radioisotope
identification using artificial intelligence.

312

P3 - 02#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

Correction of changes in Photopeak Position and
FWHM according to Temperature Changes of CsI(Tl) Detector
ChanJoong Kim1, YongHyun Kim2, HanCheol Yang2, Junehyung Bernaski1,
JooHyung Kim1, YongKyun Kim1,*
Department of Nuclear Engineering of Hanyang University / South Korea1,
Raymetrics / South Korea2
Email: ykkim4@hanyang.ac.kr
Keywords: CsI(Tl) Detector, Photopeak, FWHM
For scintillation based radiation detectors, changes in temperature cause a shift
in the photopeak position and variations in the size of the Full Width at Half Maximum (FWHM). To achieve fast and accurate identification of radioactive isotopes
and dose measurement using scintillation detectors regardless of temperaturechanges, the photopeak position and FWHM chgane correction functions were
derived by varying the detector's temperature.
The CsI(Tl) detector was used in this study and the temperature variation test
was conducted by the constant temperature and humidity chamber C4-180.
Measurements were taken for 10 minutes at each temperature with a 10°C interval.
Six check sources (Am-241, Cs-137, Ba-133, Co-60, Co-57, and Mn-54) wereused at 20℃, and for temperatures other than 20℃ (-20℃, -10℃, 0℃, 10℃,
30℃, 40℃, and 50℃), the temperature variation test was conducted using threeisotopes including Am-241, Cs-137, and Ba-133. Photopeak position and
FWHM values for each temperature were obtained through temperature test, andcalculated the relative change of peak shift and relative change of FWHM values
based on the values at 20℃.
As the temperature increases, the relative change of peak shift and FWHM values
generally tends to increase. This means that as the temperature increases from
20℃, the channel of the photopeak decreases, the pulse height increases andthe FWHM decreases. As the temperature decreases from 20℃, the channel ofthe photopeak increases, the pulse height decreases and the FWHM increases.
The obtained relative change of values can be used to derive correction functions
for the temperature-dependent changes in photopeak and FWHM.
This study aimed to derive correction functions for the changes in the photopeak
position and FWHM of the CsI(Tl) detector according to temperature variations.
The derived correction functions will be applied to the algorithm used in the CsI(Tl)detector to enable it to identify radioactive isotopes and measure radiation dose
independent of temperature changes.
313

Abstracts - Poster

P3 - 03#3 Radiation Detection and Sensor Technology

Evaluation of Output Accuracy and Precision of 3D-Printed
Scintillator (RMPS470)

Junehyung Bernaski1, Jae Chang Kim1, Yong Hyun Kim2, Han Cheol Yang2,
Yong Kyun Kim1,*
Department of Nuclear Engineering, Hanyang University/Korea1, Department of
Nuclear Engineering, Hanyang University/Korea2, Raymetrics, Seoul, Korea2
Email: ykkim4@hanyang.ac.kr
Keywords: 3D-Printer, Plastic Scintillator, Fluorescence Spectroscopy
Scintillators for radiation detection, manufactured using 3D DLP printing, arehighly versatile, allowing for rapid and precise custom manufacturing of various
shapes and sizes, making them widely applicable in nuclear reaction measurement, medical radiation measurement, and radiation detection. Therefore, theprinting output accuracy and precision of the 3D-printed plastic scintillator are
essential performance indicators that directly affect practicality and commercial
viability.
This study measured the accuracy and precision of the 3D-printed output of
RMPS470 plastic scintillator resin, developed through research by Raymetrics and
Hanyang University.
The customized DLP printer employed in this study, designed for producing
RMPS470 scintillators by Raymetrics and Hanyang University, has a minimum
pixel size of 65 μm for the x and y-axis, a 385 nm UV wavelength, and a maximum
intensity of 30 mW/cm² for the light source. The RMPS470 scintillators were produced through 3D printing after optimizing the ratio of D0241 monomer, PPO(primary dye), ADS086BE (wavelength shifter), and TPO (photoinitiator). Eachlayer was set to a thickness of 100 μm, with a total production time of 50 minutes.
Following this, 10 × 10 × 10mm³ cube-shaped RMPS470 scintillators were modeled using the 3D modeling program Autodesk Inventor. The output accuracy andprecision of the scintillators were evaluated by measuring the x, y, and z-axis
lengths using vernier calipers. Accuracy was expressed as the ratio of the sample
mean value to the true value (mean/true value × 100%), and precision was expressed as the ratio of the standard deviation of the measured values to the mean
(standard deviation/mean × 100%).
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#3 Radiation Detection and Sensor Technology

The calculated output accuracy of the x, y-axis, and z-axis was 99.34% and97.04%, and the output precision was 0.65% for the x, y-axis, and 1.13% for thez-axis, respectively. The measurement result showed that the printed RMPS470
scintillators were smaller than the initial modeling size. In particular, the accuracy
and precision along the z-axis were lower than those of the x and y-axis.

315

Abstracts - Poster

P3 - 04#3 Radiation Detection and Sensor Technology

Improvement and evaluation of anti-tumor effect by neutronc a p t u re t h e ra p y u s in g a n e w e l e m e n t — rh o d iu m b a s e d o n b e ta r a yXuan Hou1, Hironobu Yanagie2, Takehisa Matsukawa3, Daibing Yang1, Masashi
Yanagawa4, Kenji Shimazoe1, Hiroyuki Takahashi2,*
Department of Bioengineering, School of Engineering, The University of Tokyo,
Japan1, Institute of Engineering Innovation, School of Engineering, The University of Tokyo, Japan2, Department of Epidemiology and Environmental Health,Juntendo University, Japan3, Veterinary Medical Center, Obihiro University ofAgriculture and Veterinary Medicine, Japan4
Email: leo@n.t.u-tokyo.ac.jp
Keywords: neutron capture therapy, rhodium, beta ray, liposome
Neutron capture therapy (NCT) uses secondary particle to treat tumor. Boron has
been applied to NCT in clinics, and gadolinium has also attracted the attention.
However, there are some limitations, such as the short range of Auger electrons,
which are important secondary particles in gadolinium NCT. Our group attempted
a new element, rhodium, because of its advantages, such as 100% natural abundance, long range, neutron cross-section peak, and fitness to acceleratorbased neutrons.
To reduce toxicity and increase tumor accumulation, liposomes, which arespherical nanoparticles with a lipid-bilayer, were introduced and rhodium encapsulated liposomes (Rh-Lip) were synthesized following the thin film method. In
vitro cytotoxicity test was carried on and human pancreatic adenocarcinoma cellline (AsPC-1) were exposed to rhodium solution and Rh-Lip at different concentration for 24 h. Then CellTiter 96®AQueous One Solution Cell Proliferation
Assay (MTS assay) was added and absorbance was measured for cytotoxicity
evaluation. Biodistribution and anti-tumor effect was evaluated using tumorbearing mouse model. At 3 h and 6 h after administration, the mice were sacrificed. Blood, tumor, heart, lung, liver, spleen, and kidneys were harvested, andthen rhodium concentration was measured using coupled plasma mass spectrometry (ICP-MS). After administration, neutron irradiation was conducted at the
Aomori Prefecture Quantum Science Center (QSC), and the neutron fluence wasdetermined to be 1 × 10^12 n/cm^2 and 2 × 10^12 n/cm^2. After neutron irradiation, tumor inhibition ability was evaluated based on tumor growth, and tumor
size was measured every three or four days using a caliper. Tumor volume wascalculated using the formula of V= (a×b^2)∕2, where a and b are the major and316

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#3 Radiation Detection and Sensor Technology

minor axes of the tumor, respectively. Moreover, pathological analysis of eachorgan was conducted to evaluate tumor damage ability.
After 24 h exposure to rhodium solution, the cell viability increased to 90% when
the rhodium concentration was diluted to 0.25 mg/mL; in contrast, it was up to
90% when rhodium concentration was diluted to 0.063 mg/mL in the Rh-Lipgroup. Moreover, in the Rh-Lip group, 387.3 ppm rhodium remained in the tumor
3 h after administration, but only 46.6 ppm remained in the rhodium solutiongroup. After neutron irradiation, Rh-Lip with 2 × 10^12 n/cm^2 neutron irradiation showed the best tumor inhibition ability compared with other groups. Moreover, pathological analysis showed cells was damaged in Rh-Lip with irradiationgroup. However, normal histology in the Rh-Lip without irradiation group showed
healthy tumor cells without damage.
In this study, rhodium was investigated as a new element for NCT. Liposomes
have been introduced to decrease the toxicity of rhodium and increase its retention time inside tumors. Rhodium-encapsulated liposomes were successfully
synthesized by the thin film method. Based on the results of neutron irradiation,a higher tumor inhibition ability of rhodium-encapsulated liposomes was observed. Rhodium with a higher neutron fluence showed higher inhibition, which
can be considered as a low neutron cross-section. Generally, in this study, theanti-tumor effect of rhodium-encapsulated liposome was confirmed.

317

Abstracts - Poster

P3 - 05#3 Radiation Detection and Sensor Technology

Evaluation of Properties of Plastic Scintillators Fabricatedby 3D Printing

Han Cheol Yang1, Tae Hoon Kim2, Yong Hyun Kim2, Yong Kyun Kim1,*
Department of Nuclear Engineering, Hanyang University/Korea1,
Raymetrics, Korea2
Email: ykkim4@hanyang.ac.kr
Keywords: 3D printing, Plastic scintillator, Commercialization, radiation detection
Plastic scintillators are widely used in the field of radiation detection and measurement. Depending on the combination and ratio of constituent resins, the performance of plastic scintillator is varied, and the use of 3D printing for manufacturing has the advantage of being able to output various shapes. RMPS470 is aplastic scintillator fabricated using a DLP 3D printer and is currently commercialized by Raymetrics for radiation measurement. THPS is also in the commercialization stage and was set as a comparison group due to its high printing accuracy.
This study is an evaluation of the properties of RMPS470 for radiation instrument
design and medical engineering applications. The properties are composed of transmittance, density, refractive index, and EA (elemental analyzer) analysis. In addition,
THPS, which had been previously developed, was fabricated for comparison.
The transparency of RMPS470 was measured using Agilent Cary 300, whichmeasured the amount of transmitted light when light in the range of 200 nm to
800 nm was exposed. Density analysis was performed using an electronic densitometer (MD-300S, Alfa Mirage), and refractive index was measured at 404 nm,
532 nm, 632 nm, and 829 nm using the Prism Coupler (2010/M, Metricon). EA
analysis was carried out using FlashSmart and EA2000 (Thermo Fisher Scientific)
to determine the content of carbon, hydrogen, nitrogen, and oxygen elements.
Those three analyses were conducted in Koptri.
The transmittance of RMPS470 was 26.2 at 470 nm (THPS – 11.6), and the density was measured at 1.214 g/cm3 (THPS - 1.139). The refractive index was
1.642 at 470 nm, and the elemental composition was analyzed as carbon 76.85%,
oxygen 15.55%, hydrogen 5.84%, and nitrogen 0.34%. However, when producing
RMPS470 using a 3D printer to print spherical or complex models, over-curing
tendencies were observed at the bottom surface in contact with the support,
which affected the printing accuracy.
This study conducted a detailed analysis of the properties (Transparency, Density,
Refractive index, EA analysis) of RMPS which is fabricated using a DLP 3D printer.
However, in the future, measures need to be taken to prevent over-curing andimprove printing accuracy for complex modeling.
318

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P3 - 06#3 Radiation Detection and Sensor Technology

Accelerated Monte Carlo-Based Gamma-Ray Spectrometer
Modelling for Efficient Dataset Creation in Radiation Detectionand Sensor Technology

Young-su Kim1, Jeen-hyeng Kwon2, Jiseok Kim1,*
Korea Atomic Energy Research Institute, Daejeon 34057, Republic of Korea1,
Department of Nuclear Engineering, Hanyang University, Seoul 04763, Republicof Korea2, Korea Atomic Energy Research Institute, Daejeon 34057, Republic ofKorea2
Email: geesuck@kaeri.re.kr
Keywords: Gamma-ray spectroscopy, Machine learning algorithms, Monte
Carlo simulations, Geant4, Radiation detection and sensor technology
Gamma-ray spectroscopy is widely employed in various applications such as
surveying, decommissioning, and waste characterization, with an increasing
number of studies utilizing machine learning techniques. In order to create
effective gamma spectroscopy machine learning algorithms, a comprehensive
gamma-ray spectrum dataset with known radio-isotope concentrations isessential, showcasing a diverse array of concentrations to avoid overfitting.
Although Monte Carlo simulations are a dependable and cost-efficient method
for generating these datasets, they are hindered by lengthy computational times.
In this research, a gamma-ray spectrum dataset generator was developed using
Geant4 Monte Carlo simulations, significantly decreasing the time needed fordataset preparation. The generator necessitates only one simulation per radioisotope and produces spectra for arbitrary concentrations by merging singleisotope results. This method eliminates the requirement for additional simulations
for new radio-isotope concentrations, greatly reducing preparation time.
Furthermore, a spectral generation system for gamma-ray spectrometers wascreated, taking into account their inherent characteristics, such as energy
resolution, time-related properties, and distortions in the measured signals. The
model not only considers energy broadening due to energy resolution but alsocount loss and coincidence sum resulting from true or random coincidences.
Geant4 Monte Carlo simulations and an innovative methodology were used todevelop the system, which was verified through comparisons with real-world
experiments using 137Cs and 60Co radiation sources and a high-purity
germanium detector. The generated spectra exhibited strong correlations with319

Abstracts - Poster

P3 - 06

experimental spectra, and the efficiencies of major full energy peaks closely
matched experimental values. To conclude, the developed gamma-ray spectra
dataset generator and gamma-ray spectrometer modelling system facilitate therapid generation of datasets for machine learning applications while accurately
representing the intrinsic properties of gamma-ray detection systems. This
research will significantly contribute to providing reliable training data for machinelearning models in gamma spectroscopy applications within the realm of radiation
detection and sensor technology.
Acknowledgement
This work was supported by the Nuclear Safety Research Program through the
Korea Foundation Of Nuclear Safety(KoFONS) using the financial resource
granted by the Nuclear Safety and Security Commission(NSSC) of the Republicof Korea. (No. RS-2023-00240399)

320

P3 - 07

#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

Surface contamination meters and monitors based on plastic
scintillator–WLS–SiPM technology produced on SPC ‘’Doza”
Pavel Buzhan1,*, Aleksandr Karakash1
Scientific Production Company "Doza", Russian Federation1
Email: buzhan@doza.ru
Keywords: surface contamination, monitor, plastic scintillator, SiPM
The necessity to control the surface contamination of personnel and other objects is exists at all enterprises working with radioactive materials. For such control an equipment is used which most often contains either Geiger counters, or
scintillators read by vacuum photomultiplier (PMT), or proportional gas counters.
Large-area plastic scintillator with embedded wavelength shifting fiber (WLS) has
long been used in high-energy physics for ionizing radiation detection. Such an
assembly of a scintillator with WLS is traditionally called "tile". Recently, a silicon
photomultiplier (SiPM) has been widely used in high-energy physics as a photosensor for detection of the light from the WLS. The main difference of applicationof this tile in the field of surface contamination control is as follows (IEC 60325
and IEC 61098): 1) you do not need to measure the energy; 2) absence of an
external trigger; 3) such detectors should work at sun light and in wide temperature range (standards give range from +5 °C to +40 °C but at the customer's
request, the range can be extended); 4) need to register charged particles withlow energy: beta particles Emax=0.155 MeV. To fulfill the latter requirement, forabout 2 years, many experiments were made to find a suitable scintillator-WLSSiPM configuration.
We present here the result of adaptation of the described above technology forsurface contamination control equipment. The main element is the detecting
modules with a sensitive area from 146 cm2 to 465 cm2, which can register alpha,beta or gamma radiation. The choice of the area and type of particles depends
on the specific application. Final quality of the module is guaranteed by control
measurements on every step of production process: definition of the SiPM`sbreakdown voltage, test of assembly SiPM+WLS, test of the tiles, separate testof electronics, final test of complete module.
The modules have high sensitivity, large dynamic range, excellent uniformity andtemperature stability. Dynamic range of measured surface emission rate per unit
area equals: 1 – 10^6 particles/(cm^2·min) for beta-module and 0.1 – 1·10^5
particles/(cm^2·min) for alpha-module. Variations of the module response intemperature range from –20 C up to +50 C doesn`t exceed 3%. Based on a such
modules, old equipment was upgraded and a line of new ones was put into production.
321

Abstracts - Poster

P3 - 08#3 Radiation Detection and Sensor Technology

Implementation of Digital Pulse Processing for Neutron
Measurement Using Recursive Algorithms

Jae Chang Kim1, Junehyung Lee Bernaski1, Yong Kyun Kim1,*
Department of Nuclear Engineering, School of Hanyang University/Korea 1
Email: ykkim4@hanyang.ac.kr
Keywords: neutron detection, digital pulse processing, MCNP6
Estimation of special nuclear materials (SNMs) such as uranium and plutonium is
crucial for determining safeguard and disposal suitability. Various methods exist
for SNM estimation, and among them, the Differential Die-Away(DDA) method
involves irradiating the target material with external neutrons and estimating SNMs
based on the neutron measurement's time relationship. Previous research at
LANL utilized multiple 3He proportional counters and custom-developed KM200
front-end electronics to record neutron signals. However, the analog-based
KM200 module has potential issues with noise and dead time, which can affectaccurate neutron measurement times.
To address this challenge, this study implemented digital pulse processing (DPP)
for neutron measurements after the analog-to-digital converter to minimize
signal loss. We tested various filters to shape the preamplifier output signals. The
preamplifier output pulses have relatively long decay times to account for all
charges generated in the detector. In high neutron fields, this can lead to pileup effects, where new pulses occur before the decay of previous pulses iscomplete. Hence, filters that can maintain pulse amplitude while providing shorter
decay times are required.
In this study, we derived the widely-used CRn-RCm filters (n=1,2 and m=1,2,3,4)
in analog modules through recursive algorithms, determining appropriate decay
constants and time intervals for each filter. We employed suitable pole-zerocancellation algorithms to avoid undershooting in CR1-RCn filters and applied
dynamic threshold pulse detection for determining pulse amplitude.
To verify the performance of the DPP, we conducted neutron measurement
experiments using a 252Cf neutron source and 3He proportional counters. The
DPP was applied to the preamplifier output signals, and neutron counts were
determined using a threshold set at the onset of the wall effect. The resulting
neutron count rate showed an error within 3% compared to the results simulatedusing MCNP6. In the future work, we plan to apply n/gamma discrimination to thefilters used in this study and explore various other filters, such as trapezoidal and322

P3 - 08

#3 Radiation Detection and Sensor Technology

cusp filters. We aim to derive the optimal algorithm considering computation timeand accuracy, and ultimately develop a neutron measurement system for DDA
applications using Field-Programmable Gate Arrays (FPGAs) implementing thechosen algorithm.
Acknowledgement
This work was supported by the Nuclear Safety Research Program through the
Korea Foundation Of Nuclear Safety(KoFONS) using the financial resource
granted by the Nuclear Safety and Security Commission(NSSC) of the Republicof Korea. (No. 2106075)

323

Abstracts - Poster

P3 - 09#3 Radiation Detection and Sensor Technology

Coded Aperture Gamma Camera Prototype Based on CeBr3
Single Crystal Position-Sensitive and Spectrometric Detector

Yury Teverovskiy1,*, Alain Iltis2, Ghislain Zeufack2
SPC Doza, Russian Federation1
Email: yurytever@doza.ru
Keywords: Coded aperture, gamma camera, scintillator, digital sipm
Portable systems for receiving visual images of radioactive sources, called
gamma cameras, can be used for many tasks at nuclear facilities. The definitionof most wasted places can organize working procedures in such way to minimize
the dose absorbed by personnel. Gamma cameras can be mostly used at objects
with strong background radiation or in close quarters where usage of classic
dosimeters is limited and can`t give full information on radioactive sources
distribution.
Temporal Detla from Damavan Imaging is the first Compton camera available onmarket that is built on classic two position-sensitive detector technology. It uses
two CeBr3 single crystal scintillators with total volume of 26 cm3. Temporal Detla
has high registration efficiency and angular resolution, but as well as all Compton
cameras has a major drawback – it cannot visualize low energy photons. To
overcome the drawback one of the detectors is provided with coded aperture.
Coded aperture gamma-camera based on CeBr3 scintillator can provide highdetection efficiency and is possible to acquire spectrometric information. The
prototype of this camera was built. Energy resolution is 6.5% (@661 keV), spatial
resolution acquired is 1.7mm. The prototype has excellent detection efficiency:
minimal required dose for visualization is 2.5 nSv for Cs-137 isotope and 0.14
nSv for Co-57 isotope. Spectrometric information allows independentvisualization of sources with different energies.

324

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#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

Evaluation of radiation detection properties of liquid scintillators
based on modified scintillation material

Sujung Min, Sangbum Hong*, Soon Park, Changhyun Roh, Bumkyoung Seo
Department of decommissioning technology development, Korea Atomic Energy
Research Institute/Korea1
Email: sbhong@kaeri.re.kr
Keywords: Liquid scintillator, Detection, Gamma, DMC
Gamma ray and neutron detection technology is necessary for nuclear security
and non-proliferation. Gamma ray measurement is widely used for non-destructive measurement, measurement of decommissioning sites of nuclear facilities,
and radioactive waste classification. In this study, liquid scintillator that can beused not only non-destructive measurement and decommissioning site measurement fields but also nuclear security field were fabricated.
The primary solute used in this study is PPO(2.5-diphenyloxazole), and the secondary solute is POPOP(1,4-bis [5-phenyl-2-oxazole] benzene) and DMC(7Diethylamino-4-methylcoumarin). The liquid scintillator used a quartz cell with adiameter of 50 mm, and after each scintillation material was added to the solvent,it was mixed at room temperature for 2 hours or more. In order to optimize thecontent of DMC, a test was performed after fabricating a liquid scintillator withvarious concentrations. Measurement data using a liquid scintillator was compared with the results of a commercial liquid scintillator, BC501A with a diameterof 3 inches (Saint-Gobain, St. Great Lakes parkway, OHIO, USA).
The absorption wavelength range of DMC coincides with the emission wavelengthwith the primary solute like PPO and exhibits a narrower emission peak than
POPOP. The biggest advantage of DMC is that the Stokes shift is large, so theeffect on self-absorption is less than that of POPOP. Based on these characteristics, it was confirmed that DMC can be used as a secondary solute. For thecontent optimization test, gamma radiation sources (Cs-137) were used, and it
was finally confirmed that the amount of 0.04 wt% was the most optimized content. In addition, as a result of performing the neutron/gamma identification test
(PSD), it showed lower performance than the POPOP-based liquid scintillator,
but showed a difference within 10%, showing sufficient potential as a secondary
solute. Based on the results of this study, it has been proven that DMC can beused instead of POPOP. Based on this, if studies on changes in the compositionof secondary solute or the use of nanoparticles are conducted, it will be possible
to manufacture and utilize a scintillator with improved efficiency compared to theexisting scintillator.
325

Abstracts - Poster

P3 - 11#3 Radiation Detection and Sensor Technology

A study on methods to minimize pulse pile-up and dead timeerrors in high radiation fields

Sujung Min, Bumkyoung Seo, Sangbum Hong*
Department of decommissioning technology development, Korea Atomic Energy
Research Institute/Korea
Email: sbhong@kaeri.re.kr
Keywords: Pulse Pile-up, High radiation field, Correction method, Deadtime
Radioactive contamination distribution in nuclear facilities is typically measured
and analyzed using radiation sensors. However, it can be difficult to apply sensors in high radiation fields. Therefore, shielding/collimators and small-sized detectors are generally used. Nevertheless, pulse pileup and dead time correction
still a problem for high radioactivity measurements. This can lead to measurement
errors and distortion of the energy spectrum.
To address these issues, researchers have conducted signal pileup and deadtime correction studies. In this study, 2-inch LBC and 3-inch NaI(Tl) detectors
were connected to a DT5730 Digitizer (CAEN) to obtain data, and pulses were
set up to record only one pulse in a 1us window. Shield/collimator design optimization and signal correction were performed using MCNP, AI(CNN), and numerical calculation algorithm. The thickness of the shield/collimator and the angle
of incidence were optimized using MCNP. Correction algorithms and CNN wereapplied using data obtained through MCNP and experiments.
However, the correction algorithm developed and the results of applying CNN
showed that more than 98% of the signal was corrected and the spectrum wasrestored. Nevertheless, the artificial intelligence results were based on only twoor three overlapping pulse data, and the effect of noise was not considered,indicating the need for further research.
In the future, the researchers plan to conduct research using small-sized detectors (GAGG, CZT, etc.) with dimensions of approximately 10 x 10mm. Furthermore, they intend to evaluate the performance of the measurement/analysis system in an environment similar to the high radiation field of an actual nuclear facility.
In the future, we plan to conduct research using ≈10 x 10mm small size detectors(GAGG, CZT etc.). Also, the performance evaluation of the measurement/analysis system is intended to be performed in an environment similar to the highradiation field of an actual nuclear facility.

326

P3 - 12#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

Development of miniature gamma-ray imager using 3D printed
tungsten coded aperture
Joonhyuk LEE1, Changyu KO1, Minki CHAE1, Dukwoo KIM1, Seungho LEE1,
Shinhye KANG1, Youngkwang KIM1, Manhee JEONG1,*
Jeju National University1
Email: mhjeong@jejunu.ac.kr
Keywords: Miniature, Coded aperture, 3D printing, Gamma-ray imager
Most existing radiation imaging equipment is easy to carry, but there is a limit touse in narrow and limited places. These problems can be solved by improving
portability through miniaturization and weight reduction. This study is about adevelopment of the miniature gamma-ray imager using a 3D printed tungsten
mask based on a coded aperture, a GAGG(Ce) scintillator, and a silicon photomultiplier (SiPM) array.
The miniature gamma-ray imager consists of four main components. The first is
a coded aperture-based tungsten mask made with a metal 3D printer. It is in theform of a centered mosaic MURA mask pattern, and the location of the radiation
source is identified using the incident radiation, the modulation pattern generated
by it, and the decoding algorithm. The aperture size of the mask is 1.19 mm ⅹ
1.19 mm, has an aperture pitch of 1.2 mm and 10 mm thick thickness. The second is the GAGG(Ce) scintillator. This scintillator has the advantage of high effective atomic number, density, and light yield with fast decay time. It is an 11×11
array with a pixel size of 1.13 mm ⅹ 1.13 mm ⅹ 10 mm-thick and a pixel pitchof 1.2 mm. The third is SiPM array. Compared to conventional photomultipliertube (PMT), it is easy to make compact equipment. Each pixel has a size of 6
mm ⅹ 6 mm and a 4 ⅹ 4 array with a pixel pitch of 6.2 mm, and the total size
has an effective area of 25 mm ⅹ 25 mm. Finally, it consists of a signal readout
circuit that processes information obtained from the sensor.
The total size of the manufactured equipment is 38.11 ⅹ 8.11 ⅹ 138.11 mm^3,
including batteries and readout circuit boards, and uses the USB 3.0 communication protocol. After manufacturing the equipment, performance evaluations
such as angular resolution, energy resolution, filed-of-view (FOV), special resolution in full-width at half maximum (FWHM), and image sensitivity will be conducted, and the results will be announced later at the conference.
In this study, it is expected that it can be used in various ways, such as attaching
it to the work helmet for radiation workers or installing it on unmanned robots that
require small payloads such as drones. In addition, it is expected that it will bepossible to acquire images in areas where measurement is difficult, such as narrow places, through miniaturization
327

Abstracts - Poster

P3 - 13#3 Radiation Detection and Sensor Technology

Enhancing Proton Therapy Range Verification: The Multi-Slit
Prompt-Gamma Camera for Improved Conformal Dose Delivery

Youngmo Ku1, Sehoon Choi1, Jaeho Cho1, Sehyun Jang1, Jong Hwi Jeong2,
Sung Hun Kim2, Sungkoo Cho3, Chan Hyeong Kim1,*
Depaprtment of Nuclear Engineering, Hanyang University/Korea1, Proton Therapy Center, National Cancer Center/Korea2, Department of Radiation Oncology,
Samsung Medical Center, Sungkyunkwan University School of Medicine/Korea3
Email: chkim@hanyang.ac.kr
Keywords: Proton therapy, Range uncertainty, Multi-slit prompt-gamma camera,
Range measurement, Range verification
The use of proton beams in radiation therapy is an effective method for deliveringhigh conformal dose to the target due to the sharp dose falloff at the distal endof the beam. However, beam range uncertainty has hindered achieving full conformity in practice, with an average uncertainty of 4% that varies among institutions. In response to this challenge, a new range verification system prototype
has been developed, known as the multi-slit prompt-gamma camera (MSPGC).
The MSPGC is equipped with a multiple-parallel-slit tungsten collimator and 144
CsI(Tl) detectors, which provide high prompt-gamma detection sensitivity and awide field of view up to 216 mm. An automatic camera positioning system allows
for precise and rapid arrangement in various clinical situations, and an advanced
range estimation algorithm is integrated to enhance range measurement accuracy. To assess the range measurement precision of the prototype, single spot
beams with varying energies, proton quantities, and positions were tested, as
well as spot-scanning proton beams in a simulated spot scanning proton therapy(SSPT) treatment using a phantom.
Results indicate that the prototype achieved high accuracy, with a range measurement precision of
In the future, further evaluations of the prototype will be conducted in clinicalconditions with heterogeneous media, such as anthropomorphic phantoms andreal patients. Overall, the MSPGC exhibits promising potential to improve range
verification accuracy and reduce range uncertainty in proton therapy. By providing more accurate range measurements, it can help radiation therapists deliver
more precise and effective treatments, ultimately leading to better patient outcomes.

328

P3 - 14

#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

Accelerating Cargo Screening: Development and Evaluation of aHybrid Gamma Imaging System for Rapid Detection of Nuclear
and Radioactive Materials

Junyoung Lee, Goeun Lee, Taehyeon Eom, Youngmo Ku, Chan Hyeong KIM*
Department of nuclear engineering, Hanyang University, Korea
Email: chkim@hanyang.ac.kr
Keywords: Cargo screening, Hybrid gamma imaging system, Korea customs service
The Korea Customs Service is tasked with detecting nuclear and radioactive materials in imported cargo at airports and harbors. The current method of using
radiation portal monitors (RPM) and handheld survey meters lacks imaging capabilities, providing only count rates or dose rates and making it time-consuming
to localize nuclear and radioactive materials. To overcome this limitation, a highspeed gamma imaging system was developed using a hybrid gamma imaging
technique that combines Compton imaging and coded aperture imaging, which
also provides isotope identification and dose rate estimation as well as source
localization.
The high-speed gamma imaging system comprises eight rectangular (146mm×146 mm) NaI(Tl) crystals, 72 square-type PMTs, a modified uniformly redundant array (MURA) coded aperture mask, a 72-channel high-speed FPGAbased DAQ, and a notebook computer. Additionally, a dedicated motorized cart
was developed to provide high mobility, which is critical for screening a large
volume of cargo at airports and harbors.
The performance of the system was evaluated using Cs-137 check sources atvarious locations. In low dose rate (0.03 μSv/h above the background dose rate),
the system detected the source in approximately 300 milliseconds and localized
its position within approximately 2 seconds. Even in lower dose rate conditions(0.001 μSv/h incremental dose rate), the system imaged the location of thesource within 5 minutes while estimating the ambient dose rate with an error ofless than 25%.
The experiments demonstrate the ability of the gamma imaging system to quickly
and accurately locate and identify nuclear and radioactive materials in cargo. The
high-speed gamma imaging system is expected to improve the time-consuming
localization process of radioactive contaminants, given the limitation of the existing radiation detection devices at airports and harbors, which lack imaging

329

Abstracts - Poster

P3 - 14capabilities, thereby reducing the risk of transporting nuclear or radioactive materials.
Acknowledgement
This research was supported by Field-oriented Technology Development Project
for Customs Administration through National Research Foundation of Korea (NRF)
funded by the Ministry of Science & ICT and Korea Customs Service (NRF2021M3I1A1097895)

330

#3 Radiation Detection and Sensor Technology

P3 - 15#3 Radiation Detection and Sensor Technology

Addressing Temperature Variations in Outdoor Gamma Imaging:
A Novel Regression-Based PMT Gain Optimization Technique

Goeun Lee1, Junyoung Lee1, Youngmo Ku1, Chan Hyeong KIM1,*
Department of nuclear engineering, Hanyang University/Korea
Email: chkim@hanyang.ac.kr
Keywords: High-speed gamma imaging system, PMT gain optimization, temperature variations
For rapid and accurate inspection of nuclear and radioactive materials at airports
and harbor, we are developing a high-speed gamma imaging system. This system offers several advanced features, including (1) high imaging sensitivity
achieved by employing two quad-type detectors, each with four large-area
NaI(Tl) crystals (146 x 146 mm²) coupled with 36 photomultiplier tubes (PMTs),and (2) a broad energy range of imaging by utilizing a hybrid gamma imaging
technique. In hybrid imaging, high-precision estimation of gamma-ray interaction (i.e., interaction location and energy) in the crystal is required to obtain highresolution image and, in the present study, gamma-ray interaction energy andlocation are estimated by comparing the ratio of the nine PMT signals to premeasured data and summing nine PMT signals respectively, making PMT gain
maintenance critically important. One challenge faced in this system is the temperature variations during outdoor operation, which affect the uniformity of PMT
gains as the PMT gain is highly dependent on temperature. To address this challenge, we have developed a regression-based gain optimization method tomaintain the uniformity of PMT gains.
The proposed method is consisted of four steps: (1) acquiring training data (i.e.,
PMT signals) using known-energy sources, (2) calculating the gain optimization
matrix to compensate for inter-channel deviation through linear regression on
training data and true energy, (3) refining the training data and iteratively recalculating the gain optimization matrix for optimal results, and (4) applying the gain
optimization matrix to newly measured PMT signals for optimization.
To evaluate the performance of the proposed method, experiments were conducted under two conditions; maintaining original PMT gains and changing gainsof five PMTs randomly within 0.75-1.6 times range. For both conditions, the interaction estimation (i.e., position and energy) performance was compared without and with the method. Energy estimation performance was evaluated using
Cs-137 point source and energy resolution at 662 keV was compared in each331

Abstracts - Poster

P3 - 15

condition. Position estimation performance was assessed at 20 mm intervals using a Cs-137 pencil-beam source and mean displacement (MD), calculated by
the estimation error of each event, was evaluated for 49 positions in each experiment condition.
Results demonstrated that the developed method significantly improves energy
resolution irrespective of the PMT gains being maintained, with enhancementsfrom 10.67% to 6.79% when gains were maintained, and from 21.23% to 6.82%
when gains were changed. Furthermore, the developed method marginally reduced MD from 4.5±0.7 mm to 4.3±0.5 mm when PMT gains were maintained,
and when the PMT gains were changed, the MD was significantly decreased from
12.7±7.1 mm to 4.6±0.6 mm, successfully recovering the original MD via thedeveloped method. These results indicates that the developed method can effectively maintains performance even when PMT gains are affected by temperature variations during outdoor operation. Future work will incorporate backgroundradiations into the method.
Acknowledgement
This research was supported by Field-oriented Technology Development Project
for Customs Administration through National Research Foundation of Korea (NRF)
funded by the Ministry of Science & ICT and Korea Customs Service (NRF2021M3I1A1097895).

332

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#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

Development of CZT-based Portable Gamma Imaging System for
Korea Customs Service

Jae Hyeon Kim1, Jin Hyung Park1, Byeong-no Lee1, Junyoung Lee2, Kyung Min Oh1,*
Advanced Radiation Technology Institute, Korea Atomic Energy ResearchInstitute1, Department of Nuclear Engineering Hanyang University2
Email: okm1766@kaeri.re.kr
Keywords: CZT, Gamma imaging, portable
For safe customs work, radiation inspection should be done for all cargo or
goods imported into customs sites including airports and harbors. In the customs
sites, fast and precise radiation inspection is essential to improve the efficiencyof the customs work and the safety of the worker. For fast and precise radiation
inspection, we are developing a CZT-based portable gamma imaging systemthat can specify the radioisotope and localize radioactive materials in cargo andgoods imported into the customs sites. In the present study, we evaluated theperformance of the CZT-based portable gamma imaging system.
The portable gamma imaging system was composed of a single CZT detector(M400, H3D, USA) and an RGB camera (D455, Intel, USA). The CZT detector wascomposed of four CZT crystals and each crystal has a size of 22 mm × 22 mm
× 10 mm. The RGB camera has a field-of-view of 86°×57°. A dedicated housing
was also developed to equip the CZT detector and the RGB camera. The total
weight of the portable system was ~4.2 kg including the mobile battery.
To evaluate the performance of the portable gamma imaging system experimentally, 137Cs and 22Na check sources were placed at a 50 cm distance from thesystem. The measurement time was set to be 1 minute. The results showed thatthe imaging system can specify both the 137Cs and 22Na sources. The imaging
system can also image the 137Cs source and the 22Na source, respectively. It
was confirmed that the portable gamma imaging system is able to not only specify the radioisotope but also localize radioactive materials.
To enhance the performance of the system, in the future, we will equip a coded
aperture mask for low-energy gamma-ray imaging. Furthermore, an ambient
dose rate estimation functionality will be also incorporated in the system. Then,the portable gamma imaging system is expected to improve the efficiency of thecustoms work and the safety of the worker in the customs sites.

333

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P3 - 16

Acknowledgement
This research was supported by Field-oriented Technology Development Project
for Customs Administration through National Research Foundation of Korea(NRF)
funded by the Ministry of Science & ICT and Korea Customs Service(2021M3I1
A1097896).

334

#3 Radiation Detection and Sensor Technology

P3 - 17#3 Radiation Detection and Sensor Technology

Feasibility Study on Pulse Shape Discrimination of Fast/Thermal
Neutrons, and Gamma ray with a Single Detector

Chaehun Lee*, Seonkwang Yoon, Byung-Hee Won, Seong-Kyu Ahn,
Sang-bum Hong
Korea Atomic Energy Research Institute
Email: chlee80@kaeri.re.kr
Keywords: Pulse Shape Discrimination, Simultaneous Measurement of Fast/Thermal Neutrons and Gamma rays, Machine Learning
Since nuclear materials such as spent nuclear fuel emit gamma rays/neutrons ofvarious energies, and the characteristics of gamma rays/neutron emission are
different depending on the physical/chemical properties of nuclear materials. If
thermal neutrons, fast neutrons, and gamma rays are simultaneously measured
using a single detection system, it can be utilized in the fields such as nondestructive assay (NDA) and nuclear material verification for nuclear safeguards,
and radiation monitoring applications.
To measure the neutron/gamma rays simultaneously, the detector systems must
be chosen based on the characteristics of each radiation. Recent advances inthe detector technology, such as sensor material development and the Pulse
Shape Discrimination (PSD) signal processing method led to the development ofa single detector capable of detecting both neutrons and gamma rays. However,current PSD technology can only distinguish two different pulse shapes by applying different gate lengths and cannot simultaneously measure both fast andthermal neutrons, which have different interaction materials.
This study proposes a novel approach to discriminate between more than threedifferent pulse shapes, allowing for the measurement of fast and thermal neutrons and gamma rays with a single detector system consisting of three types ofscintillators and a PMT. The scintillators used for this study are CeBr3, CLLB, EJ276, and Stilbene. The pulse shapes of each scintillation output are digitized bya digitizer with 500 MS/s sampling rate and post-processed with various integration gate lengths to discriminate the interaction scintillators and radiation
types. The experimental results will be presented in the conference. In addition
to the post-processing of digitized pulse shapes, a machine learning PSD isinvestigated and some preliminary results will be discussed.

335

Abstracts - Poster

P3 - 18#3 Radiation Detection and Sensor Technology

Application of Optical Fiber Beam Loss Monitor in SSRF

Dong Zhang1, Ming Zeng2, Jianbing Zhang1, Jiaqiang Xu1,*
Department of Radiation Protection and Technical Safety, Shanghai
Synchrotron Radiation Facility, Shanghai Advanced Research Institute, China1,
Department of Engineering Physics, Tsinghua University, China2
Email: xujq@sari.ac.cn
Keywords: Beam loss monitor, Fiber, Cerenkov radiation
The Shanghai Synchrotron Radiation Facility (SSRF) was built in 2009 andprovides more than 5000 hours of experimental light per year. The device
comprises a 432 m ring that stores electrons up to 3.5 GeV, which is most brilliant
and best performance 3rd light source in middle energy interval. It is essential tounderstand the beam status in the particle accelerator, especially stored beam
as it turns around the layout or beam that is lost. Beam Loss Monitors (BLMs) inaccelerator facilities are widely used to study beam losses situation and acquire
kinds of information about environment dose for radiation safety and machine
protection. To avoid the disadvantages of SiPIN based detectors, SSRF requires
a fiber based BLM system.
Optical fibers have been proven that it can be easily to detect large area in many
accelerator facilities included linacs and storage rings. They were originallyshown as a method of efficiently protecting accelerators from beam losses
damage. This kind of system is called the optical fiber Beam Loss Monitor (oBLM).
Beam loss rates in accelerator can be achieved by modifying three beam
parameters: bunch charge, horizontal/vertical coupling and of dynamic aperture,
which are strongly correlated to the beam lifetime. The signal of beam losses isobserved by means of the Cherenkov light produced in a 1000 μm core plastic
fiber, when particles pass through the it with speed is more than light of medium.
This paper presents the process of design and preliminary test of the oBLMs,
which are incorporating into the Shanghai Synchrotron Radiation Facility as partof the upgrade proposal for current SiPIN based BLM system. The prototype will
be installed over long distance of beam lines in SSRF for radiation protection,
especially in injection area and standard straight section with undulator. A
detailed design and its commissioning of oBLMs are discussed in thiscontribution.

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#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

Time-of-Flight Neutron Bragg Edge Imaging Using a Compact
Accelerator Neutron Source

Mahdi Bakhtiari1, Hirotaka Sato2, Nam Suk Jung3, Wooyong Um1, Hee Seock
Lee3,*
Division of Advanced Nuclear Engineering, Pohang University of Science andTechnology/Korea1, Faculty of Engineering, Hokkaido University/Japan2,
Poahng Accelerator Laboratory, Pohang University of Science andTechnology/Korea3
Email: lee@postech.ac.kr
Keywords: Bragg edge imaging, Pulsed neutron source, Boron-based GEM
detector
Bragg edge imaging has become a promising technique for studying structural
materials non-destructively. The development of pixelated detectors has made it
possible to measure material characteristics in two dimensions with resorbablespatial resolution. Bragg edge imaging utilizes the neutron time of flight (TOF)
method. Neutrons are attenuated after passing through a sample and for somestructural materials such as iron, sharp peaks are appeared known as Braggedges. These edges occur at a specific neutron wavelength according to Bragg’s
law. The crystallographic information of a sample such as microstructure andtexture is deduced by analyzing the position and height of the Bragg edges.
The Bragg edge imaging was performed to measure the crystallite size, degreeof anisotropy (texture) and phase fraction of stainless-steel samples. The
samples included one cold-rolled, 5 annealed samples at 950, 1080, 1160, and1200 °C, and one hot-rolled. The dimensions of the cold-rolled and annealed
sample were 2.5×2.5×0.14 cm3 and that of hot-rolled was 4×4×0.3 cm3.
Neutron transmission was measured using a boron-based gas electron multiplier(GEM) detector with a detection area of 10×10 cm2 ¬(0.8 mm pixel size).
Irradiation times with and without samples were 13.4 and 10 h, respectively. A
grid collimator (Gd2O3) was attached in front of the detector to remove neutrons
scattered by the samples. All the samples were then placed in front of thecollimator and irradiated simultaneously. The experiment was performed at
Hokkaido University Neutron Source (HUNS) providing a 2.3 kW electron beam
power and neutron wavelength resolution of 0.5% at 0.4 nm at a distance of 6.24
m from the neutron production target. Instrument resolution was measured using
an α-Fe texture-free sample and applied in the analysis. Neutron transmission

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spectra were analyzed with the Rietveld Imaging of Transmission Spectra (RITS)
code using extinction and March-Dollase functions for crystallite size and texture
analysis.
The study revealed that the steel samples crystallite size increased with theannealing temperature. Cold-rolled and annealed samples exhibited strong
textures, while the hot-rolled sample showed no texture. These results wereverified by electron backscattered diffraction (EBSD) which is a conventional
method of material characterization.
Bragg edge imaging was performed without destructing the bulk samples. This
work also confirmed the ability of small-scale neutron sources to be used for thisimaging technique.

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#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

A machine learning-based peak area calculation algorithm in low
background measurement

Peng Zhang1
Key Laboratory of Particle and Radiation Imaging (Ministry of Education) and
Department of Engineering Physics, Tsinghua University, Beĳing 100084,
China1
Email:
Keywords: Peak area,Machine learning,Low background measurements
In radioactive measurements, peak area analysis of the energy spectrum isneeded to determine radioactivity. Traditional methods, including direct andfunction fitting-based methods, heavily rely on assumptions made regarding thebackground, which render them impracticable in the absence of accurate
knowledge of the background. Moreover, as the radioactivity approaches the detection limit, traditional methods incur greater uncertainty in calculation. In order
to meet the data analysis requirements of low background measurements andenhance detection sensitivity, we propose a machine learning-based method forpeak area analysis.
Initially, we acquired a diverse spectrum database of low background corresponding to distinct background scenarios using the Monte Carlo sampling technique. We then utilized this database to train and validate the model. Following a
5-fold cross-validation analysis of the performance of different machine learningalgorithms, we conducted rigorous parameter tuning to better train the model. In
order to assess the efficacy of the model under diverse conditions, we calculated
forecast errors using a completely independent test set.
After successfully validating the model using the test set, we observed more
precise and reliable results in comparison to traditional methods. Additionally,
this approach exhibits a significant reduction of dependence on background assumptions and can produce superior predictions under varying background conditions.
Finally, we applied our method to analyze low background radioactivity spectrum
recorded by the GeTHU spectrometer at China JinPing Underground Laboratory(CJPL), and compared the results obtained using our method with those obtainedusing traditional methods. We noted a comparatively reduced degree of uncertainty in the calculation results, thereby enhancing detection sensitivity of low
background spectrometers.

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P3 - 21#3 Radiation Detection and Sensor Technology

Experimental verification of real-time gamma-ray energy
spectrum and dose monitor up to 3 MeV

Nikolaos Voulgaris1, Hayato Sumi1, Sachie Kusaka1, Tamaki Shingo1,
Isao Murata1,*
Division of Sustainable Energy and Environmental Engineering, Graduate Schoolof Engineering, Osaka University, Osaka, Japan1
Email: murata@see.eng.osaka-u.ac.jp
Keywords: Radiation detection, dose rate, gamma-rays, BNCT, Bayesian estimation
PurposeOur research group has been developing a portable spectrometer, that can simultaneously measure the energy spectrum and dose of gamma-rays in real time.
The main focus of this project is medical applications and the aim is increasing
the awareness of medical professionals about radiation exposure (1). The basic
characteristics of the monitor have been verified and measurements in the background and in front of a nuclear fuel storage room, with a spectrum similar tothat of a radiological treatment room, have been performed. The results showedthat spectral reconstruction and dose estimation can be carried out simultaneously and accurately (2). Confirmation of the true real-time estimation method
has also been conducted, verifying the applicability of the monitor to estimate
the gamma-ray spectrum and dose rate in true real-time (3). However, only
measurements of gamma-rays with energies up to 1.3 MeV have been conducted.
Thus, the purpose of this research is to verify that this prototype monitor canmeasure gamma-rays with energies of up to 3 MeV.
Methods
The monitor consists of a CsI (Tl) crystal (2.6 x 2.6 x 2.6 cm3), a multi-pixel
photon counter (MPPC), an amplifier and until recently, a multi-channel analyzer(MCA8000D), while currently the DP5 digital pulse processor is being used. The
pulse height spectrum is converted into an energy spectrum by the unfolding
process, with the response function. The dose is calculated by applying a doseconversion coefficient to the energy spectrum. From the quantitative relationshipbetween the measured pulse height spectrum and the gamma-ray energy theenergy spectrum can be estimated using the Bayesian estimation, as describedby Iwasaki (4). The application of the sequential Bayesian estimation in radiation
measurement states that the energy spectrum can be inferred, iteratively as the340

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#3 Radiation Detection and Sensor Technology

pulse-height data is received for each count by the detector. Thus, in order torealize accurate real-time estimation, an improved sequential Bayesian estimation method is used. In this study, we confirm experimentally that the prototype
gamma-ray detector can detect gamma-rays of up to around 3 MeV. For thispurpose, an Al foil was irradiated with DT neutrons in order to induce the27Al(n,α)24Na reaction, which produces radioactive 24Na, emitting gamma-raysof 1.368 MeV and 2.754 MeV. Following irradiation, measurements were performed using a Ge detector, an NaI survey meter, and the prototype real-timemonitor.
Results
The gamma-rays were measured by the prototype detector and the energy peaks
were accurately detected and observed in real-time. The results of the dose rate
were compared with the results of the NaI survey meter. Due to the calibration of
the NaI survey meter with a 137Cs (0.662 MeV) source, an expected discrepancy
with the results of the prototype monitor was observed. Finally the theoreticalvalue of the dose rate is calculated and compared with the estimated experimental results of the real-time measurements of the dose rate, which was foundto be around 4 μSv/h at 20cm.
Conclusions
From the results we conclude that the monitor can display the energy spectrum
and dose-rate in true real-time in an environment of up to 2.754 MeV gammarays. Currently, extension of application to 10 MeV is under investigation.
Reference
1) Kobayashi, M., Sato, F., Kusaka, S., & Murata, I. EPJ Web of Conferences,
153, 07014. (2017) 2) Nishimura, H., Shinohara, M., Miyoshi, T., Voulgaris, N.,
Kusaka, S., Tamaki, S., Sato, F., & Murata, I. Applied Radiation and Isotopes,
110226 (2022) 3) Voulgaris, N., Miyoshi, T., Tamaki, S., Kusaka, S., Murata, I.,
yBNCT 2022, November 14-18, (2022) Abstract submitted for publication 4) Iwasaki, S., Proceedings of the 9th International Symposium on Reactor Dosimetry:
Prague, Czech Republic, 2-6 September 1996. Singapore: World Scientific, p.
245–252. (1998)

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P3 - 22#3 Radiation Detection and Sensor Technology

Remote Radiation Detection Imaging System Mounted on a Drone

Youngjoo Park1, Inbum Lee1, Kangwoo Jeon2, Donghui Seo1, Kyunghoon Cho2,
Jaewon Jeong2, Hakjae Lee2, Kisung Lee1,*
Department of Biomedical Engineering, School of Korea University/Korea1,
ARALE Laboratory Co., Ltd/Korea2
Email: kisung@korea.ac.kr
Keywords: radiation detection system, radiation monitoring, unmanned aerial
systems, gamma-ray imaging, gamma camera
Radiation disasters occur worldwide. Their damage and scale are increasing. In
the case of radiation disasters caused by natural disasters, it is difficult toinvestigate the exposure amount by approaching close to the accident site. In
addition, there is a possibility of unexpected excessive radiation exposure if
investigators do approach, which could result in unnecessary additional
casualties. To address this situation, we developed an unmanned aerial vehicle(UAV) based radiation detection system that could measure radiation disasters inreal-time from a remote location.
This system consisted of three main modules. The first was small but highperformance gamma-ray detector module mounted on a drone. The second was
a communication module responsible for data communication of drones withgamma detector and controllers. Finally, the monitoring module could provide
convenience in operation by performing fusion of optical and gamma images andautomatic energy histogram analysis. As an experimental equipment, DJI’s
Matrice 300 RTK drone was used. A small gamma detector was built on
collaboration with the ARALE Laboratory. An android-based tablet (CPU1.8GHz+1.6GHz, Octa-Core, 8 inch) was used for monitoring. Main parametersof the collimator that could affect the performance were determined using Monte
Carlo simulation software (Gate v9.1). The collimator thickness was 8 mm. The
scintillator thickness was 7 mm. The distance between the collimator and thescintillator was 20 mm. The collimator was made of tungsten and the scintillator
was made of GaGG:Ce.
To evaluate performance of the developed small detector, when the distance
between the source and the detector was 2 m, indoor and outdoor experiments
were conducted to measure each resolution of 7 and 10 degrees, respectively.

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#3 Radiation Detection and Sensor Technology

The indoor experiment confirmed that angular-resolution of 7 degrees came out
for Cs-137 of 1 mCi. In this case, the FWHM was 2 pixels and the distance was
24 cm. The outdoor experiment by hovering a drone confirmed that angularresolution of 10 degrees came out for Ba-133 of 250 μCi. In this case, the FWHM
was 3 pixels and the distance was 35 cm. However, some radiation was detected
in neighboring pixels of the detector because it tended to shake under the
influence of wind during hovering.
The goal of this study was to develop a radiation detection system using a drone
equipped with a small gamma detector, and to develop a high-performance small
gamma detector and monitoring application. As a result of the research, a small
yet high-performance detector was developed using a Coded Aperture Collimator,and data transmission was successfully achieved using the drone's own
communication system. In addition, intuitive and convenient monitoring ispossible with automatic energy analysis and optical and gamma image fusion
functions. In the future, we aim to develop a stable system by incorporating aradiation source identification algorithm and advancing software.
Reference[1] Yuki Sato, et. al, “Remote radiation imaging system using a compact
gamma-ray imager mounted on a multicopter drone“, Journal of Nuclear Scienceand Technology, Vol 55, 2018. [2] Stephen R. Gottesman and E. E. Fenimore,"New family of binary arrays for coded aperture imaging," Appl. Opt. 28, 43444352 (1989) [3] Caroli, E., Stephen, J.B., Di Cocco, G. et al. Coded aperture
imaging in X- and gamma-ray astronomy. Space Sci Rev 45, 349–403 (1987).
https://doi.org/10.1007/BF00171998 [4] Jerry L. Prince & Jonathan M. Links,
Medical Imaging Signals and Systems, 2nd Ed., Pearson, 2015 [5] ARALE
Laboratory,Maetel16specification,http://www.aralelab.com/content/detector.pdf

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Development of a Coded Aperture Collimator for Drone-Based
Radiation Imaging System

Inbum Lee1, Youngjoo Park1, Kangwoo Jeon2, Donghui Seo1, Kyunghoon Cho2,
Jaewon Jeong2, Hakjae Lee2, Kisung Lee1,*
Department of Biomedical Engineering, Korea University, Seoul, Korea1,
ARALE Laboratory Co., Ltd, Korea2
Email: kisung@korea.ac.kr
Keywords: gamma camera, coded aperture, drone
To prevent radiation accidents and minimize the spread of accidents during disasters, quick initial response is crucial. However, the current process of deploying personnel might result in casualties due to radiation exposure. Therefore, adrone-based radiation imaging system that can detect exposed radioactive isotopes in real-time from a remote location is necessary. However, the payload
weight that can be carried by a drone is very limited. It is usually only a few
kilograms. In this study, we developed a lightweight and sensitive coded aperturecollimator for gamma-ray detectors mounted in a drone.
The pattern used was a mosaicked modified uniformly redundant array (MURA).
Due to auto-correlation of this pattern with decoding matrix being a delta function, it is possible to perfectly restore the signal without any errors in the decoding
process. In this study, we designed it to have a field of view of 45 degrees andan angular resolution of 7 degrees or less. We used a coded aperture with a rankof 19 to achieve this performance. To derive appropriate performance parameters
for the detector, simulations were conducted for 4 mm, 6 mm, and 8 mm collimators. Ba-133 (0.02Ci) and Cs-137 (0.02Ci) sources were used to evaluate
imaging capability for both low and high energy isotopes. After considering bothweight and image quality, the suitable collimator thickness was determined to be
6 mm and 8 mm, which were found to be effective for imaging in terms of bothpeak to signal noise ratio(PSNR) and normalized mean squared error(NMSE)
These results were verified by producing actual collimators and conducting further testing.
Based on simulation results, experiments were conducted to compare the performance of collimator for Ba-133 (250uCi) and Cs-137 (1mCi) radiation sources.
In the case of Ba-133, comparing 6 mm and 8 mm thickness collimators, the useof an 8 mm-thick collimator resulted in a better full width half maximum(FWHM)
of 2.3 mm than the 6 mm thickness collimator. Additionally, using the 8 mm344

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#3 Radiation Detection and Sensor Technology

thickness collimator, high-resolution gamma images with a 7-degree angularresolution could be obtained when imaging Ba-133 and Cs-137.
In this study, a collimator with coded aperture was developed to create a lightweight gamma camera.
Reference[1] S. R. Gottesman and E. E. Fenimore, New family of binary arrays for coded
aperture imaging, Appl. Opt. 28, 4344 (1989)

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P3 - 24#3 Radiation Detection and Sensor Technology

Simulation of boron-coated honeycomb tube neutron detector
with high detection efficiency

Jianjin Zhou1,*, Hengguan Yi1, Sanqiang Xia1, Hengchun Zhao1, Qinjian Cao1
China Institute for Radiation Protection1, Shanxi Provincial Key Laboratory for
Radiation Safety and Protection2, CNNC Key Laboratory for Radiation Protection
Technology3
Email: zhoujianjin@outlook.com
Keywords: Neutron detection, High detection efficiency, Boron-coatedhoneycomb tube, He-3 alternative
For decades, 3He-filled proportional counters have been widely used for theneutron detection in a series of applications, such as homeland security, neutronscience and special nuclear materials search, due to their high detection
efficiency and excellent n/γ discrimination. As the supply of 3He gas decreases
and demand increases, there is an urgent need for alternatives to neutron
detection. The boron-coated straw tube neutron detector (BCS) is an attractive
3He replacement solution. A boron-coated honeycomb tube neutron detection(BCH) is proposed, and higher detection efficiency is obtained by increasing theboron-coated area than BCS.
A BCH is composed of 7 regular hexagonal tubes with petals and round corners,
these tubes with wall thickness 50 μm and side length 4.42 mm are sealed in a
regular hexagonal Al tube with wall thickness 1 mm and side length 13.92 mm.
The petal is half the length of the regular hexagon, which is used to increase theboron-coated area. The round corners radius is 0.5 mm, which is used to
optimize the electric field of the detector. The inner wall of tubes is coated witha thin layer of 10B-enriched (abundance: 96%) boron carbide (10B4C), which
relies on the 10B(n, α)7Li reaction to detect neutrons. The tubes are filled withan Ar/CO2 mixture (90/10) at 1.0 atm. The anode wire is located at the center ofthe regular hexagonal tubes.
A detailed Geant4 simulation study of the BCH is presented, which investigates
the thermal neutron detection efficiency and the deposition energy of secondary
ions with the different 10B4C thickness. As the thickness of 10B4C increases,
the detection efficiency first increases to the maximum and then decreases. The
detection energy threshold is set to 20 keV. For a single BCH, the optimum346

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#3 Radiation Detection and Sensor Technology

thickness of 10B4C is 1.3 μm, and the corresponding detection efficiency is
24.7%. For the 3-layer, 5-layer and 7-layer BCH, the optimal thickness of 10B4C
is 0.9 μm, 0.8 μm, 0.7 μm, respectively, and the corresponding detection
efficiency is 46.9%, 55.8%, 61.0%, respectively.
BCH is a neutron detector with excellent performance, which can achieve highdetection efficiency comparable to 3He-filled proportional counters throughmulti-layer stacking.

347

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P3 - 25#3 Radiation Detection and Sensor Technology

Characteristic Determination of 1,800 cc Ion Chamber and Radiation Measurement for Human Body Scanner
Yong Ho Jin1, Min Woo Kwak1, Hyun Woo Kim1, Kwang Pyo Kim1,*
Kyung Hee University, Yongin, Korea1
Email: kpkim@khu.ac.kr
Keywords: Human Body Scanner, Ion Chamber, ANSIAs the number of drug smuggling in Korea continues to increase, the importanceof human body scanning at airports and ports to prevent smuggling has become
increasingly important. Human body scanners used at airports and ports use xrays, millimeter waves, and terahertz waves. Among them, human body scannerusing x-ray causes radiation exposure to examinee. Therefore, in order to reduce
the radiation dose to the examinee, manufacturers of human body scanner are
applying the latest imaging technology (e.g., fan beam irradiation). To measure
the fan type beam with conventional detector might be difficult. Therefore, theAmerican national standards institute (ANSI) has provided guidance on the selection of the appropriate detector for measuring body scanner. This guidance is:first, the device must be able to adequately measure the cumulative dose. Second, it should have high sensitivity, and third, the guidance suggests using alarge, unpressurised ion chamber which has large volume (at least 1,500 cc).
Therefore, in this study, 1,800 cc ion chamber was used. The 1,800cc ion chamber has a large volume with a height of 171 mm and a diameter of 138 mm.
Therefore, when radiation is irradiated from the human body scanner, there is a
disadvantage that the radiation enters with “partial irradiate”. In this reason, thecharacteristics of the 1,800cc ion chamber were determined before measuring
the human body scanner. Especially, fan-type radiation was focused on thisstudy. As the result of the characteristic determination, the dose differences between “partial irradiate” and “full irradiate” was found to be insignificant. Therefore, 1,800cc ion chamber can be used to measure human body scanner. The
irradiation scope mode is divided into torso/full body/both modes, and the dosemode is divided into High/medium/low. In this study, the irradiation scope mode
was set as torso/full body/both mode each, with “high” dose mode. The dose of
torso/full body/both irradiation scope mode was 0.9, 1.0, 1.9 μGy/scan each.
This study can be used as a basis for the development of techniques for radiation
measurement and safety assessment for body scanners.
Acknowledgement
This work was supported by a grant from “Development of techniques for radiation measurement and safety assessment for body scanners” carried out by KINS.
348

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#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

Neutron multiplication and prompt neutron angular distributions ofdiverse nuclear material samples

Seonkwang Yoon1,*, Chaehun Lee2, Young-Su Kim2, Hee Seo2, Ho-Dong Kim2
Advanced Fuel Cycle Technology Development Division, KAERI, Korea1,
Department of Quantum System Engineering, JNU, Korea4, Advanced Fuel Cycle Technology Development Division, KAERI, Korea5
Email: skyoon94@kaeri.re.kr
Keywords: Nuclear Safeguards, fast neutron measurement, anisotropic emission,
neutron multiplication, nuclear verification
The angularly biased emission properties of prompt fission neutrons have beenstudied in diverse perspectives in terms of nuclear physics and safeguards. In
such preceding studies, it was confirmed that neutron multiplication, which is a
critical unknown to deter 240Pu effective mass of subcritical multiplying nuclear
material samples, could be estimated through direct measurement of fast neutrons. Also, the nuclides-dependent characteristics of anisotropic emission ofprompt neutrons were shown according to energy thresholds for diverse sample
masses.
However, the practical measurement of the anisotropic emission properties would
be interfered by diverse geometric variables including sample mass, composition,and density as well. Most of the previous results focused on the changes ofmeasurable anisotropic parameter, that is asymmetry of prompt neutron angular
distribution, with respect to sample masses of pure metal samples but, indeed,
one of the significant concerns is the density or volume of targeted samples. It
was found that the loss of primary directional information of fission neutrons bymultiple scattering within a sample volume affects the results of measured anisotropic property, as much as even comparably to neutron multiplication by fission chain reactions that is corresponding to sample mass change. To verify,
therefore, the multiplication and asymmetry tendencies as sample mass changes
at different density is necessary prior to the feasible investigation.
The material states are not constant and could be varied as the purpose of nuclear facilities and applications. To that end, in this work, several nuclear material
sample types were assumed, for instant the specified plutonium isotope(s) ofpure metal, metal oxide (PuO2), and monazite (Pu(PO4)) whose densities areabout 20, 10, and 5, respectively. The first type can be seen in electrochemicallyextracted samples from spent nuclear fuels (SNFs), and the second type can be349

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seen in normal SNFs, and the last one can be seen in stabilized solidified materials of SNFs. Also, several samples was supposed to be containing 244Cm lessthan 1wt% compared to plutonium.
Acknowledgement
This work was supported by the National Research Foundation of Korea (NRF)
grant funded by the Ministry of Science and ICT of Republic of Korea (2021M2E3
A3040093 and RS-2022-00154985).

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#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

Development of a New Low-Energy Neutron Spectrometer Usinga Position-Sensitive Proportional Counter -Examination of
Response Functions Including Shield-

Yu Fujiwara1, Masahiro Hori1, Shingo Tamaki1, Sachie Kusaka1, Fuminobu Sato1,
Isao Murata1,*
Osaka University1
Email: murata@see.eng.osaka-u.ac.jp
Keywords: Boron Neutron Capture Therapy, Accelerator-Based Neutron Source
BNCT, Neutron Detection, Spectrum Unfolding, Response Function
PurposeOur study aims to establish a method to accurately measure the energy spectrumof low-energy neutrons used in Boron Neutron Capture Therapy (BNCT) [1].
BNCT can kill cancer cells selectively, so it's expected to apply to refractory
cancers such as Glioblastoma[2]. Currently, Accelerator-Based Neutron Source
BNCTs (ABNS-BNCT) are under development to be utilized as the neutron source
instead of the nuclear reactor. Therefore, evaluating the characterization of theradiation field in ABNS becomes a crucial issue. However, it is difficult because
their features depend on the kind of accelerator or type of moderator. In particular,
the “energy spectrum of the epi-thermal neutron” is too difficult to find a method
to measure correctly[3]. Our laboratory has been developing a new type of spectrometer based on a 3He position-sensitive proportional counter. In this device,signals can be obtained from both ends and the position where neutrons reactwith 3He can also be obtained by making a ratio of both signals [4]. We defined
the depth of neutron detection position from the entrance side of the device as“Detection depth (r).” The measured detection depth distribution (Y(r)) can beconverted to the energy spectrum (φ(E)) by solving an inverse problem with theestimated detector response function (R (E, r)), Y = R φ. To solve this inverse
problem, Bayesian Estimation is applied. To validate this method, several experiments have been carried out so far [5,6]. The procedure is as follows; First, asa neutron source, a thermal/epi-thermal column with a DT neutron source of
OKTAVIAN in Osaka University was developed. Next, validation is conducted by
the comparison between the estimated value of the φ1(E) from the measured Y(r)and the simulated value of the energy spectrum at the exit of the designed thermal/epi-thermal column (φ2(E)). However, there have been discrepancies be-351

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tween φ1(E) and φ2(E) [6]. In this presentation, I show the new method to improve them.
Methods
We focus on the response function R (E, r). In the previous research, it wasdetermined theoretically based on the detection position of neutrons from theentrance of this device. In this case, a beam neutron source is necessary for thevalidation experiment, because if neutrons enter from the side of the device, theevents can give errors in the results of Y(r). Then, in the present study, the spectrometer was redefined to include the shield surrounding the device, and R (E, r)
was evaluated by Monte Carlo code for the redefined spectrometer. It has been
confirmed that MCNP allows the accurate creation of response functions [7]. In
this study, the distance between the epi-thermal column and the spectrometer
was varied, and R was calculated for each case.
Results
As a result, we found that the response functions computed have a different
shape compared with those in the previous studies. In addition, we have finallysucceeded in designing the experimental system to validate this method.
Conclusion
In this study, we developed a method to calculate and apply the detector response using Monte Carlo simulation in order to further improve the accuracy ofthe measurement method of the energy spectrum in the low-energy neutron region using position-sensitive proportional counters. Furthermore, it was confirmed that the values of the response functions calculated by the simulationdiffered from the theoretical values, and we completed the design of the system
to validate this method. We plan to conduct validation experiments to verify thatthe newly obtained value (φ3(E)) and φ2(E) are consistent with each other.
Reference[1] Locher, G, L. Am.J.Roentgenol.Radium Ther. 1936, 36, 1-13. [2] Wolfgang
A.G. Sauerwein, Andrea Wittig Raymond Moss, Yoshinobu Nakagawa Editors;“Neutron Capture Therapy” Springer. [3] Murata, I., et al. "Techniques to measure absolute neutron spectrum and intensity for accelerator based neutron sourcefor BNCT," Plasma and Fusion Research 13, (2018). [4] W. N. McElroy, et al., A
Computer Automated Iteration Method for Neutron Flux Spectra Determination,
report AFWL-TR- 67-4 1, Air Force Weapons Laboratory (1967). [5] ] I, Murata,
H.Miyamaru: "Low-energy neutron spectrometer using position sensitive proportional counter -Feasibility study based on numerical analysis -" Nuclear Instruments and Methods in Physics Research A 589,445-454(2008). [6] Y. Osawa,et.al., “Development of An Epi-thermal Neutron Field for Fundamental Researches for BNCT with A DT Neutron Source,” EDP Sciences (2017). [7] M.
Kobayashi, Master-dissertation, Experimental Possibilities for a Real-Time γ-Ray
Spectrum and Dose Simultaneous Measurement System, 2016352

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The performance analysis of the laundry monitor

Jing Zhang1,*, Yi Ren1, Xiangyang Du1, Jianfei Wang2, Xirong Guo1
China Institute for Radiation Protection1, Shanxi Zhongfu Nuclear Instruments
Co.Ltd2
Email: zhangjing@cirnic.com
Keywords: Laundry monitor;Physical performance of detector;the monitoring ofβ radioactive contamination
Based on the regulations related HAF nuclear safety and the standards related
HAD, combined with years of the nuclear power on-site operation and maintenance as well as the experience feedback, a laundry monitor is designed. The
physical performance, electrical safety regulations, electromagnetic compatibility
and environmental adaptability of the instrument are tested according to relevant
standards, and the test results meet the corresponding requirements. The technical qualifications and the human-computer interaction meet the design requirements. The measurement process is fully automated, which can effectively
monitor and sort the surface radiation contamination of clothing, and prevent the
spread of radiation contamination.

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P3 - 29#3 Radiation Detection and Sensor Technology

Investigation of a Statistical Spatial Analysis Methodology for
Measurement-based Target Point’s Spatial Dose Rate Estimation.

Ji Woo Kim, Hee Won Lee, Min Seong Kim, Kwang Pyo Kim*
Dpartment of Nuclear Engineering, Kyung Hee University/Korea1
Email: kpkim@khu.ac.kr
Keywords: Dose estimation, Kriging interpolation, Inverse distance weighted
Emergency workers may be used to prevent further damage in an emergency at
NPP. To optimize the dose of emergency workers, it is necessary to establish awork plan through radiation dose assessment. However, there may be a lack of
information for dose assessment in an emergency, such as the source term.
Therefore, a methodology for radiation dose prediction based on measurement
information that is easier to obtain is required. To improve the accuracy of
predicting worker dose based on measurement information, it is necessary to
determine the spatial dose rate along the worker's path. However, obtaining theexact target point's measurement information in every situation is impossible.
Therefore, a methodology for predicting the spatial dose rate at the specifictarget point is needed. In this study, we investigate the methodologies that canbe applied to derive the spatial dose rate at a specific point from measurements.
For this purpose, statistical spatial analysis methodologies such as the kriging
and IDW(Inverse Distance Weighted) methods were investigated. Kriging
probabilistically predicts the value of the target point by reflecting the linear
combination of the measured points and the correlation between the measured
values. As one of the representative spatial interpolation techniques, it canestimate the predicted value, variance, standard error, etc., together, and can be
useful for interpolating contours and surfaces when the observation points are
irregular. However, it requires much computation and can be more timeconsuming than other methodologies. IDW is a deterministic method thatinterpolates weights based on the distance from the data. Compared to the
kriging method, it is relatively fast and easy to calculate. However, the maximum
and minimum values may exist only at the measure points due to the nature of
the interpolation method. Furthermore, the quality of the interpolation results may
be poor if the data distribution is uneven. In this study, we investigated themethodology applicable to estimating spatial dose rates at a specific point frommeasured values. Statistical spatial analysis methodologies such as kriging and
IDW can be used for spatial dose rate estimation. The results of this study can354

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be used as a basis for target point’s spatial dose rate estimation based on
measurements.
Acknowledgement* This work was supported through the National Research Foundation of Korea(NRF) using the financial resource granted by the Ministry of Science and ICT(MSIT). (No. RS-2022-00143994).

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P3 - 30#3 Radiation Detection and Sensor Technology

Development of Online Directional Dose Equivalent Rate
Monitoring System

yingjing wei1, liye liu2, zhifang wu1
Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing ,China1, China Institute for radiation protection , Taiyuan, China；2
Email:
Keywords: weak penetrating radiation; online radiation monitoring; H ̇(0.07);
plastic scintillator; SiPM
β Radiation widely exists in many fields, such as nuclear industry, radiation
medicine, radioisotope applications, and so on. It is more likely to cause local
high dose radiation to organs , such as human limb skin and eye lenses, and is
not easily detected. For radiation protection safety, research and development
online β Radiation monitoring system, real-time monitoring of sites β Radiationdose changes are particularly important to provide protection for the radiation
safety of personnel in the workplace.
In order to solve the problem of measuring high and weak penetrating radiationin the field, a ZF-FD7 online H´(0.07) monitoring system was developed basedon plastic scintillators and SiPM.
Referring to the test methods given in IEC60846.1:2009, the radiation
characteristics of the developed ZF-FD7 online H´(0.07) monitoring system were
tested. The test results are as follows: in the range of 76μSv/h ~ 1.4Sv/h, its
relative inherent error is within ±10%; The average energy of the β ray is 60 keV~ 800 keV, and the response is within the range of 0.8~1.2; At 0 ~ ± 45º
incidence, the response of the incidence angle is between 0.95~1.09; The
repeatability of the measurement results in a 1 mSv/h radiation field is 2.0%. Aftertemperature compensation, the response change is 0.91~1.08 within -10°C ~+40°C, and 0.93~1 in the range of relative humidity of 45% ~ 85%. The test
results show that the characteristics of the ZF-FD7 online H´(0.07) monitoring
system meet the requirements of IEC60846.1:2009.
This online H' (0.07) Radiation monitoring system can solve the problem of highand weak penetration radiation. The Radiation monitoring system can be widelyused in nuclear facilities in the future. The monitoring data of H'(0.07) will provide
support for radiation protection decisions.

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#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

Design and simulation verification of a coded aperture gamma
camera with different vertical and horizontal fields of view

Yan Li1, Hengguan Yi1,*, Sanqiang Xia1, Chongyang Wang1, Xiaolong Wang1
Shanxi Provincial Key Laboratory for Radiation Safety and Protection, CNNC Key
Laboratory for Radiation Protection Technology, China Institute of radiation
protection,China1
Email: kidyhg1412@sina.com
Keywords: M-M coding mode; Monte Carlo simulation; Gamma camera;
Gamma ray detection; Field of view
[Background]Currently, more mainstream coding methods, such as Uniform
Redundant Arrays (URA) or Modified Uniform Redundant Arrays (MURA), requirethat the vertical and horizontal field of view be approximately the same or thesame. However, in some practical applications, it is often the case that thevertical field of view is small enough to fully include the object. The larger thehorizontal field of view, the better for the entire system, For example, themeasurement of the source term position of pipelines in nuclear power plants.
Moreover, neither of these coding methods is self-supporting and requires
additional mechanical structures for fixation, which makes it difficult tomanufacture during active alignment. [Purpose] In order to solve the problem
that existing coded aperture gamma cameras based on URA or MURA encoding
methods must have the same vertical and horizontal fields of view and cannot be
self-supported, a M-M encoding method with self-supporting capabilities, which
can have different vertical and horizontal coding sequences, is proposed to
produce a coding collimator for coded aperture gamma cameras. [Methods] The
method of the imaging system based on the M-M coded method is discussed,then the model of the imaging system based on the M-M coding method isestablished by the Monte Carlo software MCNP. The encoding array of the entire
imaging system is the 25 × 33 array of 10mm × 10mm × 4mm pixels, andthickness requires more detailed simulation. The material of the detector is GAGG
(Ce) crystal, the array was designed as a 13 × 17 array of 10mm × 10mm ×10mm pixels. The distance between the coded collimator and the detector arrayis 14.5 cm, with a horizontal field of view of approximately 56 °, and a verticalfield of view of approximately 42 °. [Results] Using radiation sources of different
energies at different positions to irradiate coding collimators with different
thicknesses, it is determined that the optimal thickness of the coded collimator

357

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P3 - 31is 10mm by analyzing and explaining the reconstructed image SNR obtained.
Then the response of the imaging system based on the optimal parameters ofcoded mask to the single-point source, multi-point source and extended source
was simulated. The reconstructed image SNR of the single-point source reached40.36, and the imaging system also successfully reconstructed the shape of theextended source. There are not fluctuating sidelobes in the background ofreconstructed images of point and extended source. [Conclusion] The above
results prove the excellent performance of the M-M coded method, and it has
the potential to replace MURA or URA in special scenarios. This article can lay
the foundation for the subsequent development of coded aperture gamma
camera based on M-M coding.

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#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

Application of new UAV radiation aerial survey technology innuclear emergency

mei feng1,*
Department of Health Physics, China Institute for Radiation Protection, China1
Email: fengmei_cirp@126.com
Keywords: unmanned aerial vehicle, radiation aerial survey technology, nuclear
emergency
With the development of nuclear energy and nuclear industry, higher
requirements are put forward for the response to emergency scenarios such as
nuclear anti-terrorism and radiation accidents. Unmanned Aerial Vehicle (UAV)
equipped with radiation detectors and necessary functional devices is aneffective method for rapid and efficient location of pollution hot spots and searchof uncontrolled radioactive sources, which can provide necessary informations
for the decision-making of radiation protection(MA Xiaoyu, 2018).
In this paper, a measuring system is designed and built based on a small multirotor UAV. Because of it's low flight altitude, the system is even more sensitive
than conventional fixed-wing UAV measurement systems for radiation detection,and can perform remote control, real-time measurement, safety analysis, andrapid response, it is especially suitable for the location of pollution hot spots andrapid source-seeking operation in most radiation emergency scenarios.
The system's hardware mainly consists of two parts, a UAV and a airborne pod
system . The UAV chosen the general model of DJI T40, which can carry 20
kilograms and fly for 20 minutes. The airborne pod is the key components of thesystem, which needs to carry functional loads such as radiation detector, position
locator, range finder and wireless communication module. It must satisfy thefunctions of high-speed flight, wide range measurement, high analysis sensitivity,
real-time satellite positioning, precise detection range and wireless datatransmission.
The airborne pod system's carry loads are mainly as follows:
A NaI(Tl) radiation detector with a dimension of 76mm in diameter and 76 mm inheight, it's large crystal structure can realize a high sensitivity detection a wide
range counter of HNC-001.
A satellite positioning module of BN-808 GPS.
A detection distance measurement module of B-200.
A communication module of AS69-T20.

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The pulse width modulation (PWM) duty cycle method is used in the system's
data analyze algorithm. Which can compensate the photomultiplier tube's errorthat occurs when the count rates is saturated. It greatly improves the accuracyof the survey results.
Tests were carried out in the open air. Two cesium sources with activities of
3.61MBq and 667MBq respectively, one cobalt source with an activity of 279MBq
were arranged in a triangle with different spacing distances. The UAV flied at analtitude of 1.5 meters and a speed of 1 meter per second, with a scan interval of
1.5 meters and a count taken every 0.25 seconds. The location of the three
radioactive sources can be clearly identified after a scan, and their activities canbe expressed in different colors.
Reference
MA Xiaoyu, MENG De. (2018) Nuclear Electronics ＆ Detection Technology, Vol.
34, 409-413.

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#3 Radiation Detection and Sensor Technology#3 Radiation Detection and Sensor Technology

Monte Carlo Simulation for An Underground Radioactivity
Measurement

Chen Jikai1
Department of Engineering Physics, Tsinghua University, China 1
Email:
Keywords: rare-event search experiments;low-background; uBq/kg; high purity
germanium
Low-background gamma-ray spectrometrys based on high-purity germanium(HPGe) detectors are employed for material screening and selection in rareevent search experiments in astroparticle physics. The materials used in rareevent search experiments have low activity at the level of mBq/kg even μBq/kg.
GeMPI3 is the most sensitive gamma spectrometer with a background count ratesof 24±1 counts/kg/day in the energy range of 60-2730keV among low-background gamma spectrometers, while Gator has a background count rates of
89±0.7 counts/kg/day in the energy range of 100-2730keV. An underground radioactivity measurement system would be operated in China Jinping Underground
Laboratory(CJPL). The system would be employed for the material screening andselection tasks for CDEX and Panda. Other experiments can also benefit fromthis technology.
The system is equipped with 15 gamma-ray low-background germanium detectors. Each HPGe detector is proposed to have a germanium crystal with a relative
efficiency of 100%. The yields of radionuclides other than 68Ge are calculated
from the beginning of the crystal growth. 68Ge is almost near the saturation. The
shield includes a 1 m layer of polyethylene, a 15 cm layer of common lead, anda 15 cm layer of high purity copper. 210Pb, a subsidiary of the 238U decay chain,is the biggest contributor to the background at a level of roughly 60% with onlya 5 cm layer of high purity copper according to the experience of GeMPI. A
contamination of cosmic radionuclide such as 60Co and 58Co is to be expectedin copper. The lead-copper shielding structure is arranged in the detector roomwith the polyethylene covering the walls and floor. The copper sample is thecylinder shape with a size of and a mass of 2.4 kg in the simulation. The known
background components is modeled using Monte Carlo simulations. The framework used for the Monte Carlo simulations is based on Geant4(currently version
10.06). The precise simulation is operated in the detection efficiency calculation

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and the background model. The shield and the detector arrangement are optimized with the help of simulations.
The background is a mix of factors including the radon in the sample cavity, thecosmogenic nuclides in germanium, the structures inside the detector, the passive shield, and the external background from the rocks and concretes. The Environmental background can be effectively shielded with the passive shield. The
most significant background is from the cosmogenic nuclides in germanium and
222Rn in the sample cavity. The second biggest contributors are the copper
shield and copper structures inside the detector. The lead shield background can
not be ignored. The background counts rates can be reduced to 18 counts/kg/
day in the energy region of 100 keV to 2730 keV with methods of backgroundcontrol. The minimum detectable activity (MDA) for a 2.4 kg sample is 0.56mBq/
kg for 232Th decay chain with 30 days measure time. MDA could achieve thelevel of 100 μBq/kg with larger sample with a mass of 100 kg.
The performance of gamma spectrometers is mainly affected by the detection
efficiency and background count rates. MDA could achieve the level of 100
μBq/kg with larger sample. The optimized design can meet the need of CJPL for
low-background measurements. The commissioning phase is planned to start in
2024 in CJPL to ensure performance.
Reference[1]Biancacci Valentina. The GERDA Experiment in the Search for Neutrinoless
Double-Beta Decay[J]. Moscow University Physics Bulletin,2022,77(2). [2]Papoulias D. K.,Kosmas T. S.. Standard and Nonstandard Neutrino-Nucleus Reactions Cross Sections and Event Rates to Neutrino Detection Experiments[J]. Advances in High Energy Physics,2015,2015. [3]E. Aprile,K. Arisaka,F. Arneodo,etal. Material screening and selection for XENON100[J]. Astroparticle Physics,2011,35(2). [4]W. Maneschg,M. Laubenstein,D. Budjáš,et al. Measurementsof extremely low radioactivity levels in stainless steel for GERDA. Nucl. Instrum.
Meth. A593 (3), 448–453. [5]Qian Zhicheng,Si Lin,Abdukerim Abdusalam,et al.
Low radioactive material screening and background control for the PandaX-4T
experiment[J]. Journal of High Energy Physics,2022,2022(6). [6]G.Heusser,M.
Laubenstein,H. Neder,et al. Low-level germanium gamma-ray spectrometry atthe μBq/kg level and future developments towards higher sensitivity[J]. Radioactivity in the Environment,2006,8. [7]Ackermann Nicola,Bonet Hannes,Buck
Christian,at el. Monte Carlo simulation of background components in low level
Germanium spectrometry[J]. Applied Radiation and Isotopes,2023,194. [8]Araujo Gabriela R.,Baudis Laura,Biondi Yanina,et al. The upgraded low-background germanium counting facility Gator for high-sensitivity γ-ray spectrometry[J]. Journal of Instrumentation,2022,17(08). [9]Ramírez García D.,Baur D.,
Grigat J.,et al. GeMSE: a low-background facility for gamma-spectrometry at362

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moderate rock overburden[J]. Journal of Instrumentation,2022,17(04). [10]Zhi
Zeng,Yuhao Mi,Hao Ma,et al. The characteristics of a low background germanium
gamma ray spectrometer at China JinPing underground laboratory[J]. Applied
Radiation and Isotopes,2014,91. [11]Akerib D. S.,Akerlof C. W.,Akimov D. Yu.,etal. The LUX-ZEPLIN (LZ) radioactivity and cleanliness control programs[J]. The
European Physical Journal C,2020,80(11). [12]V. Ã lvarez,I. Bandac,A. Bettini,et al. Radiopurity control in the NEXT-100 double beta decay experiment[J].
AIP Conference Proceedings,2013,1549(1). [13]Ramírez García D.,Baur D.,
Grigat J.,et al.GeMSE: a low-background facility for gamma-spectrometry at
moderate rock overburden[J]. Journal of Instrumentation, 2022,17(04). [14]V.B.
Brudanin,V.G. Egorov,R. Hodák,et al.Development of the ultra-low background
HPGe spectrometer OBELIX at Modane underground laboratory[J]. Journal of
Instrumentation,2017,12(2). [15]Xingyu Pan,Hao Ma,Zhi Zeng,et al. Optimal design for a μBq/kg gamma spectrometer based on Monte Carlo simulation[J].
Applied Radiation and Isotopes,2020,157(C). [16]Thiesse M.,Scovell P.,Thompson L.,et al. Background shielding by dense samples in low-level gamma spectrometry[J]. Applied Radiation and Isotopes,2022,188. [17]Knoll, G.F., 2015.
Radiation Detection and Measurement. [18]G.Sýkora,M. Ješkovský,R. Janik,et al.
Low-level single and coincidence gamma-ray spectrometry[J]. Journal of Radioanalytical and Nuclear Chemistry,2008,276(3) [19]Arpesella, C., Back, H.O.,
Balata, M., Beau, T., Bellini, G., Benziger, J., et al., 2002. Measurements of extremely low radioactivity levels in BOREXINO. Astropart. Phys. 18 (1), 1-25. [20]Baginova M.,Vojtyla P.,Povinec P.P.,et al. The effect of neutrons on the background of HPGe detectors operating deep underground[J]. Astroparticle Physics,2022,143. [21]JingLu Ma,Qian Yue,ShinTed Lin,et al. Study on cosmogenic
activation in germanium detectors for future tonne-scale CDEX experiment[J].
Science China Physics, Mechanics & Astronomy,2019,62(1). [22] M.Agostini,A.
M. Bakalyarov,M. Balata,et al. Modeling of GERDA Phase II data[J]. Journal ofHigh Energy Physics,2020,2020(3). [23]Róbert Breier,Jozef Masarik,Veronika
Palušová,et al. Monte-Carlo calculation of production rates of cosmogenic radionuclides in a HPGe detector operating in the Modane underground laboratory[J]. Nuclear Inst. and Methods in Physics Research, A,2020,978. [24]Li
Wang,Qian Yue,KeJun Kang,et al. First results on 76 Ge neutrinoless double beta
decay from CDEX-1 experiment[J]. Science China Physics, Mechanics & Astronomy,2017,60(7). [25]Leonard D.S.,Hahn K.I.,Kang W.G.,et al. Developmentof an array of fourteen HPGe detectors having 70% relative efficiency each[J].
Nuclear Inst. and Methods in Physics Research, A,2021,989. [26]Erica Andreotti.
Characterization of BEGe detectors in the HADES underground laboratory[J].
Nuclear Inst. and Methods in Physics Research, A,2013,718(C).

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P3 - 34#3 Radiation Detection and Sensor Technology

3D radiation mapping: 3D scene and gamma radiation dose ratemeasurement data fusion

Hui LI1, Qing FAN1, Hua LI1, Faguo CHEN1, Liye LIU1,*, Sanqiang XIA2, Yuan
ZHAO3
China Institute for Radiation Protection, Taiyuan, China1, Shanxi Provincial Key
Laboratory for Radiation Safety and Protection, Taiyuan, China2, Key Laboratoryof Radiation Protection Technology of China National Nuclear Corporation,
Taiyuan, China3
Email: liuliye@cirp.org.cn
Keywords: 3D radiation mapping, multi-sensor data fusion,3D ALARA digital
twins
Comprehensive, accurate, and precise radiological and topographic data are
required for the planning and implementation of safe and efficient operation,
maintenance, decontamination, and decommissioning projects of nuclear
facilities. However, the traditional radiation mapping approach hardly adapts toefficiently tracking changing radiation fields, heavily depends on labor andmanagement intensively data collection methods, and is not compatible with 3DALARA digital twins system. Therefore, intelligent 3D radiation mapping
instruments need to be developed to solve these problems.
This paper focuses on the development of an intelligent gamma radiation dose
rate field 3D mapping device. The hardware of this device is composed of a
gamma dose rate meter, an RGBD camera, an inertial measurement unit, and adata processing unit. Multi-sensor data fusion technics have been used in thisdevice. Visual SLAM(simultaneous localization and mapping) algorithm was used
to map the 3D indoor environment of the nuclear power plant. Gamma dose rate
data was acquired during the mapping of the environment and the dataacquisition time and space location matched with the mapping data
simultaneously. Consequently, the 3D environment and gamma dose rate
distribution can automatically format a 3D radiation dose field with 3D spaceinterpolation algorithms. This information can be further utilized to infer radiation
hot spots and assist decision-making for radiation protection schemes.
The device is modularized, as the radiation detector can be replaced. The gamma
dose rate meter developed by our CIRNIC team was used. It can measure H*(10)range 0.01 μSv/h~10Sv/h, its measurement error was ±10%(0.2mSv/h, 137Cs),
and its energy response was ±30% in 50keV~1.5 MeV. An office with 137Cs

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#3 Radiation Detection and Sensor Technology

source on the desktop was mapped with the instrument we developed, and a 3D
point cloud of office model was built with dose rate data and path data in it. The
data information of the 3D model and dose rate is going to infer the location ofradiation hot spots. The trajectory accuracy will be explored. The software of thedevice plan to be upgraded and different radiation detectors and different robot
mobile platforms will be configured in the future.
Acknowledgement
This study was supported by the Shanxi Science and Technology Innovation
Talent Team and China National Nuclear Corporation Elite Talent Project.

365

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P3 - 35#3 Radiation Detection and Sensor Technology

Research status and development of neutron detector
hengchun zhao1
China Institute for Radiation Protection1, Shanxi Key Laboratory for Radiation
Safety and Protection2
Email:
Keywords: neutron detector,research Status,development trend
The measurement of neutrons is important for the aerospace industry, nuclear
security, industrial production, basic scientific research, and radiation medicine.
Commonly used neutron detectors include scintillator detectors, semiconductor
detectors, gas ionization detectors, thermoluminescence detectors, and selfsufficient energy detectors. Many countries have conducted in-depth researchon these neutron detectors.
Scintillation detections are one of the most widely used ionizing radiation detectors. The scintillators used for neutron detection include organic scintillators andinorganic scintillators. Organic scintillators have high neutron detection efficiency
and fast time response, while inorganic scintillators have the characteristics ofsmall size and good scintillation performance.
The semiconductor detector uses semiconductor material as the detection medium. The detection principle of the semiconductor detector is that charged particles enter the sensitive volume of the semiconductor detector to generate electron-hole pairs, and the electron-hole pairs drift under the action of an external
electric field to output signals. Semiconductor detectors have the characteristics
of high energy resolution and wide linear range.
Gas ionization detectors refer to radiation detection that uses gas as the detection medium. The main principle is that the ray ionizes the working gas in thedetector, and then collects the generated electric charge to detect the ray. Gas
neutron detectors have the advantages of simple manufacture, strong radiation
resistance and high detection efficiency.
Thermoluminescence dosimeters use the principle of thermoluminescence torecord radiation doses. Its advantages are small size, good portability and low
price. The disadvantage is that the reading cannot be repeated, the lower limit ofthe dose that can be measured is high and the uncertainty is large.
Self-sufficient energy detectors are mainly used for the measurement of neutron
flux rates in nuclear reactors. They have the advantages of simple structure,strong radiation resistance and long service life.
With the development of science and technology, neutron detectors will face
more high-energy physics, strong radiation fields and space environments in the
future. Future neutron detectors should be portable, economical and able toadapt to the harsh environment of space.
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P4 - 01

#4 Environmental Radiation Measurement and Assessment

Natural radioactivity in bottled water and associated age-dependent
dose and lifetime risk assessment in Korea.
Ji Eun LEE, Hyo Jin KIM, Yong Uk KYE, Hee Jin JANG, Yoon Ha KIM, Wol Soon
JO, Chang Geun LEE, Jeung Kee KIM, and Yeong-Rok KANG*
Dongnam Institute of Radiological & Medical Sciences, Seoul, Korea

Monitoring drinking water, including bottled water, is imperative to safeguarding
public health, especially where bottled water consumption is high like in Korea.
The world Health Organization(WHO) has issued guidelines for drinking water
quality. If the radioactivity dose from drinking water is below the WHO recommended guidance level of 0.1 mSv/y, water is deemed safe for human consumption from a radiological perspective. Some studies investigating the radioactivity
levels in bottled water have reported wide ranges in activity concentration of naturally occurring radionuclides. In this study, focuses on the activity concentrations of 40K, 226Ra, 232Th and 3H in bottled water available in the Korea market, toassess the radiological impact to different age groups.
A High Purity Germanium(HPGe) detector (GC4018, Canberra Inc., USA) wereused for 40K, 226Ra, and 232Th analysis of the samples. Each water sample wasmeasured at 80,000s. For the determination of 3H, the spiked duplicate method
was employed. 8ml of each water sample was taken for each radionuclide andmixed with 12ml of Ultima GoldTM LLT(PerkinElmer Inc., USA) liquid scintillation
cocktail. The measurement samples was repeatedly measured 10 times for 30
minutes each.
Activity levels for all radionuclides were below the allowable maximum guideline
values specified by local and international standards. The average radioactivity
concentrations of 40K, 226Ra, and 232Th were 704, 72, 110 mBq/L. The guidancelevels value calculated from the measured radioactivity concentration shows alower value than other countries including WHO. The recorded 3H activity levels
for all investigated drinks of waters are below the 100-10,000 Bq/L specified by
EU, WHO, United States etc. guidelines for drinking water. The annual effective
doses were calculated for various age groups as well as male and female within
the population associated with each radionuclide investigated. Calculated agedependent intakes indicated that infant, children, adult groups received the highest doses for the 40K, 226Ra, and 232Th radionuclides, respectively. Adult males
exhibited a higher ELCR for three isotopes, compared to females. Based on thisit can concluded that bottled water in the Korea is safe from the radiologicalaspect of the investigated radionuclides, and does not pose any radiologicalhealth risk to the public.

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P4 - 02#4 Environmental Radiation Measurement and Assessment

Optimization of the detection system for real-time monitoring ofmarine radioactivity

Mee Jang1,*, Hyun Chul Kim2, Min Sun Lee3, Young Hak Kim4, Jae Hyeon Seo5,
Jong Myoung Lim6
Environmental Radioactivity Assessment Team, Korea Atomic Energy ResearchInstitute/Korea1, Environmental Radioactivity Assessment Team, Korea Atomic
Energy Research Institute/Korea2, Environmental Radioactivity Assessment
Team, Korea Atomic Energy Research Institute/Korea3, Environmental Radioactivity Assessment Team, Korea Atomic Energy Research Institute/Korea4, Environmental Radioactivity Assessment Team, Korea Atomic Energy Research Institute/Korea5, Environmental Radioactivity Assessment Team, Korea Atomic
Energy Research Institute/Korea6
Email: mjang@kaeri.re.kr
Keywords: Minimum Detectable Activity, Marine radioactivity, Real-time monitoring
Due to the Fukushima polluted water discharge plan and the increasing numberof nuclear power plants in East Asia, the public's interest and anxiety about marine radiation environment is increasing. When using the USV (Unmanned Surface
Vehicle), it is possible to move the accident area more actively than a ship witha fixed route in general, and to perform an exploration according to the planned
coordinates. In previous study, to confirm the possibility of monitoring marine
radiation, mobile radiation detection technology was tested at sea. As exploration
results, a similar level of background radiation was measured and we evaluatedthe MDA of the system. However, since in-situ monitoring has a higher minimum
detectable activity (MDA) than laboratory analysis, it is necessary to improve the
MDA for reliable monitoring results.
Factors affecting the MDA are measurement time, sample amount, detection efficiency and background. When we use the mobile platform for monitoring ofmarine radioactivity, the detection efficiency can be lower than laboratory analysis due to the detector size and sample size. And usually the measurement timeis also very short, the MDA is very high compared to laboratory analysis.
In this research, we designed and simulated the detection system using theMonte Carlo code to improve the detection efficiency. After optimization of thedetector system, the MDA of the system can be lower than previous study. The
results can be used to increase the reliability of exploration results, and can beapplied to various planforms and used throughout the entire life cycle of a nuclear
accident.
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#4 Environmental Radiation Measurement and Assessment#4 Environmental Radiation Measurement and Assessment

A study on the method of measuring environmental level 137Cs
radioactivity in seawaterDong Hye Heo1, Eun Hee Kwon2, S.H. Hwang, B.C. Kim, K.B. Lee, J.M. Lee*
Ionizing Radiation Metrolgy Group/Division of Chemical and Biological Metrology/Korea Research Institute of Standards and Science (KRISS)/Korea1, University of Science and Technology (UST)/korea2

Email: jmlee@kriss.re.kr
Keywords: 137Cs , seawater, AMP co-precipitation method, direct measurement
method,gamma-ray spectroscopy

The strong social demand for environmental radiation monitoring in Korea, which
was triggered by the 2011 Fukushima nuclear power plant accident in Japan, hasrecently been further aggravated by the discharge of contaminated water. And
the Republic of Korea stipulates that the minimum detectable activity (MDA) for
each radionuclide must be observed when monitoring environmental radiation.
The MDA of the specified gamma-emitting nuclide 137Cs is 3 mBq/kg, which is
not of a level that can be easily satisfied.
In Korea, organizations in charge of monitoring radioactivity in seawater usuallyconcentrate about 60 kg of seawater by the AMP co-precipitation method andcount it using gamma-ray spectroscopy. It is known that the total measurement
time required to obtain a measurement result through a concentration process
for one type of sample is at least 3 days.
Recently, a gamma ray spectrometer having relatively high efficiency and improved stability required for long-term spectral accumulation has been widelyused. Using these conditions, constructing a measurement geometry that canincrease efficiency and counting for a sufficient time to satisfy the MDA will enable more efficient monitoring of environmental radiation.
In this study, we tried to compare the characteristics and measurement resultsof each measurement method by comparing the conventional measurement
method using the AMP co-precipitation method and the direct measurement
method in which the detection efficiency was increased and the spectral accumulation time was increased without special pretreatment. Seawater samples forcomparison of measurement methods were prepared and used with artificial sea-369

Abstracts - Poster

P4 - 03water to which 137Cs was added at an appropriate level. In the direct measurement method, two high-purity Ge-detectors with relative efficiencies of 30 % and
70 % were used.

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#4 Environmental Radiation Measurement and Assessment#4 Environmental Radiation Measurement and Assessment

One-Class SVM Based Anomaly Detection for LSTM-Based
Radiation Prediction

sihyun lee1, raehyun lee*Department of Electronic Engineering1
Email: fogus1998@naver.com
Keywords: LSTM, One-Class SVM, radiation Prediction
This study emphasizes the importance of identifying pollutants and their effects
on our environment through environmental monitoring. Environmental monitoring
aims to continuously monitor various environmental factors such as air, water
quality, and radiation, and to prevent and solve environmental problems bycollecting and analyzing information on environmental changes.
In this study, we confirmed that a multivariate data model including both radiationdose rate and weather data is required for radiation dose rate prediction. To thisend, LSTM was used for seasonal data prediction, and the performance of themodel was improved by removing abnormal values of dose rate using one-classSVM.
The One-Class SVM learns using only one class data, which sets the dose rate
to one class and finds the hyperplane. The hyperplane aims to maximize the area
where dose rate data is located and to maximize the margin with other data.
Anomaly detection model learns with dose rate data only and models the
distribution of dose rate data using one-class SVM. This model is used to classify
new data, and data located on one side of the hyperplane is determined to be
outliers. Therefore, by detecting and removing outlier data for dose rates, theperformance of the radiation dose prediction model can be improved.
After detecting and removing outliers using the One-Class SVM, the radiationdose rate was predicted using the LSTM. A model that uses input data andprevious states to predict the next state. You can consider the temporaldependence of the input data by remembering the information that occurred in
the previous state and considering it in the next state. This was to predict
seasonal radiation dose rate data.
Finally, the performance of the radiation dose rate prediction model using oneclass SVM and LSTM was evaluated. The performance of the One-Class SVM
and LSTM models was evaluated through various metrics such as MSE and371

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P4 - 04

confusion matrix, accuracy, precision, reproduction rate, and F1 score, and theperformance of the One-Class SVM was 89.87%, accuracy of 1.0, reproduction
rate of 0.8987, and F1 score of 0.9467. The MSE of the LSTM was 0.00156,
which showed a lower value than that of the model without the One-Class SVM.
Therefore, One-Class SVM performed well in removing outliers from radiationdose rate data through outlier detection, and LSTM performed well in predicting
seasonal radiation dose rate data considering time dependence. Therefore, thisstudy demonstrated that the radiation dose rate prediction model combining oneclass SVM and LSTM has high performance.
Through this study, we developed a dose rate prediction model that combines
one-class SVM and LSTM using real-time ambient radiation dose and weather
information data. By leveraging one-class SVMs that improve predictive
performance by detecting and eliminating outliers and LSTM that predicts
seasonal data considering time dependence, we demonstrate that efficient dose
rate prediction is possible. These findings provide important information toprevent and resolve the various problems that arise from environmental
monitoring, and in the future, they can contribute to preventing and solving
environmental problems by actively utilizing predictive models in this way.

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P4 - 05

#4 Environmental Radiation Measurement and Assessment#4 Environmental Radiation Measurement and Assessment

An Approach to Calculation of Individual Doses on Explosion Accident Scenarios during Decommissioning of Nuclear Power
PlantsKwangho Jo1, Hyu-Chang Choi, Jung Hyejin, Ryu Gangwoo
Decommissioning Technology Group, KHNP CRI, KOREA1
Email:
Keywords: decommissioning accident, explosion accident, individual dose
After the nuclear power plant is permanently shut downed, it is different from theaccident type considered at the operation stage. One of the most difference indecommissioning is the occurrence of accidental criticality is not envisaged in
shutdown nuclear reactors from which the fuel elements have been completely
removed. However, given that there is radioacitivity that can be emitted off-site
if the existing confinement barriers are damaged by an accident of malfuntion, it
is necessary to evaluate the individual dose for accidents that might occur duringdecommissioning activities.
The accident are conservatively approach that all radioactivity emitted to the offsite is released without considering any filters. The main exposure pathways are(1) external exposure by radioactive clouds from gamma and beta ray, and (2)internal exposure by suction of contaminated air. And the methodology for evaluating the impact of public due to radiological accident is the same as in operation phase. In addition, the atmospheric diffusion factor (X/Q) is applied at the
Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) for 2 hours, andconservatively, the emission type can be applied to the ground release. However,
the methodology for defining released amount is differnt approach from the operation phase.
According the NUREG-0586 document, the types of accidents that may occur
during the decommissioning activities can be broadly classified into drop, fire,
explosion, loss of confinement, liquid spill. In this paper, an explosion accident
during the RV cutting activity was analyzed. Using the RV source term presented
in IAEA SRS No.95, the amount of radioactivity released by cutting and the individual dose was calculated.
Reference
IAEA SRS No.95, Methodology for Assessing the Induced Activation Source Termfor Use in Decommissioning Applications.
373

Abstracts - Poster

P4 - 06#4 Environmental Radiation Measurement and Assessment

LSC method for determination of Total-Uranium and Uranium
Isotopes in Liquid Effluents from nuclear fuel cycle facilities

Yang Youkun1,*, Bao li1, Guo chen1, Yang Hailan1
China Institute for Radiation Protection, Taiyuan, China1
Email: 1216981116@qq.com
Keywords: LSC, total-uranium, liquid effluent
Extraction liquid scintillation method is an analytical method suitable for rapid
measurement of total uranium and uranium isotopes in liquid effluents from nuclear fuel cycle facilities. By means of extraction and separation, it is possible tobetter remove impurity nuclides (especially α And the liquid scintillation spectrometer α The detection efficiency of nuclides is close to 100%, which has significant advantages compared to other measurement methods; Liquid scintillation
spectrometer α/β The screening function can also achieve the measurement of
uranium isotopes. By selecting the best PSA value, the spectral peaks of uranium
isotopes can be distinguished from the energy spectrum, thereby determining thebest measurement channel. The extraction liquid scintillation method can also be
combined with other measurement methods to further confirm the reliability ofthe experimental method.
As the main nuclide in nuclear fuel cycle facilities is uranium, monitoring of uranium and uranium isotopes can establish a correlation analysis between the content of effluent and production and operating conditions. Monitoring of liquid
effluent can reflect the operating conditions of facilities involved in liquid effluent
such as production workshops and wastewater treatment facilities in real time,
which is conducive to timely finding and solving problems.
The liquid scintillation method is used to analyze the liquid effluents from thefront-end facilities of the nuclear fuel cycle and measure the important contents
of total uranium and uranium isotopes in the liquid effluents α The activity concentration of radionuclides can improve the monitoring of liquid effluents. At thesame time, monitoring data can be used to feedback the operating conditions of
the facility, facilitating the monitoring of operating conditions through monitoringdata.
Based on the liquid scintillation method, the liquid scintillation method can be
combined with other measurement methods to further develop more complete
and mature measurement methods. While ensuring the accuracy of measurement
data, the measurement progress can be improved, the detection lower limit canbe reduced, and the analysis process can be simplified.
374

#4 Environmental Radiation Measurement and Assessment

P4 - 07#4 Environmental Radiation Measurement and Assessment

Main Performance Test of Gaseous Effluent Sampling System

Li Qiao1,*, Yi Ren, Haiyu Hong, Jianfei Wang, Yi Zhang
China Institute for Radiation Protection, Taiyuan Shanxi, China1
Email: qiaoli@cirnic.com
Keywords: gaseous effluents; aerosol; iodine; tritium; carbon; noble gas;
sampling and measurement
In order to solve the problems of the sampling system of gaseous effluents innuclear power plants, such as separation arrangement of monitoring equipment,difficult sampling operation and unstable sampling, etc. The self-developed
gaseous effluent sampling system can reasonably place the aerosol iodine
sampler, tritium sampler, carbon sampler, hygrometer, pressure gauge and mass
flowmeter. The intelligent display of the temperature, humidity, flow rate, pressureand other information of the sampled gas on the industrial computer is convenient
for personnel to operate and view. Different login passwords are added to prevent
various misoperation, and intelligent shutdown of pump is added to better
increase the security of the instrument and operators. Through the main
performance test, the results show that the capture efficiency of aerosol is≥99.9% , the capture efficiency of elemental iodine is ≥95% , the capture
efficiency of organic iodine is ≥80% , the the capture efficiency of H-3 including
tritium water (T2O, HTO) and tritium gas (HT, T2) was ≥97% , and the the capture
efficiency of C-14 including C-14 in CO2 and C-14 in CO、CH4、C2H6、C3H8
was ≥97% . The gaseous effluent sampling system is now undergoing long-term
stability tests and is performing well.
Acknowledgement
Topic number: 4 ( Environmental radiation measurement and assessment)
Presentation: Poster Student competition: No Telephone: +86-15234125230 Fax:
+86-15234125230 e-mail: qiaoli@cirnic.com；715585182@qq.com

375

Abstracts - Poster

P4 - 08#4 Environmental Radiation Measurement and Assessment

Gamma nuclides discrimination technology based on fully connected neural network

Xin Liu1, Ri Zhao1,*, Na Liu1, Jing Zhang1, Runcheng Liang1, Zhaoxing Liu1
China Institute for Radiation Protection, China1
Email: zhaor.abc@163.com
Keywords: Fully Connected Neural Network, Monte Carlo, Nuclides Discrimination
NaI detector has become the most widely used detector by virtue of its low priceand high detection efficiency. However, the low energy resolution of this type ofdetector leads to the low accuracy of the results of traditional nuclide discrimination methods. In recent years, with the development of artificial intelligenceand machine learning, the accuracy and speed of nuclide discrimination havemade a qualitative breakthrough. However, many methods directly use existing
deep learning models to analyze gamma spectrum, which lack interpretability. To
solve this problem, an interpretable full spectrum discrimination algorithm basedon fully connected neural network is proposed. The full spectrum discrimination
method takes each count of the gamma spectrum as the input value of the neuralnetwork, which can make full use of the spectral data, so as to improve theaccuracy of the results.
The data set used in the training was simulated by FLUKA, a Monte Carlo simulation software. In order to match the simulation spectrum with the experimental
spectrum as much as possible, and achieve the effect of combining online training and offline discrimination, we conducted a detailed modeling of NaI detector,
and used the Gaussian broadening coefficient and energy calibration curve determined according to the experiment during the simulation. The scenarios of
three gamma sources (137Cs, 60Co, 54Mn) at different distances and combinations were simulated.
The results show that the accuracy of FLUKA simulation data is more than 90%
for the mixed spectrum composed of 1~3 gamma sources in different scenarios.
In order to verify the effectiveness of the proposed method, experiments are also
carried out on the real radioactive source data, and the experimental results show
the feasibility and effectiveness of the proposed method. With the help of Monte
Carlo simulation data, the proposed method avoids disadvantages such as insufficient experimental conditions and repetitive and boring acquisition of training
data, and realizes the discrimination and interpretation of measured data.

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P4 - 09

#4 Environmental Radiation Measurement and Assessment#4 Environmental Radiation Measurement and Assessment

Research on computational efficiency optimization algorithm
based on source weight discrete
Wei Jiafu1, Li Hua1,*, Zhao Yuan1, Jin Chenghe1, Song Peitao1, Cao Qinjian1,
Long Zeyu1, Wang Yu1
China/China Institute for Radiation Protection1
Email: 1157759266@qq.com
Keywords: 3D radiation field , point-kernel integration,source weight discrete,efficiency,calculation accuracy
3D radiation field simulation technology can intuitively display the source term of
the nuclear facility site and the strength and change of radiation field. In order totimely get the changes of the field radiation field with the source term, it is necessary to improve the calculation efficiency of the current radiation field.
The point-kernel integration method is a relatively efficient method to calculate
the radiation field. When using point-kernel integration to calculate the radiation
field, the more discrete number of source terms, the more accurate the radiation
field calculation result will be. When the number of discretes increases to a certain extent, the calculation error tends to 0. Further increasing the number of
discretes will not improve the calculation accuracy, but will greatly increase thetime required for calculation. Meanwhile, with the increase of the distance between the calculation point and the source, the number of discretes required
when the calculation value becomes stable will gradually decrease. The weight
discrete method is adopted, that is, for the calculation points in the region closer
to the source item, a larger number of discrete points is adopted. For the calculation points far from the source term, a smaller discrete number is used.
For the line source and cube source, the calculation points with different distancefrom the source are studied, and different source discrete methods are used tocompare the error of the calculation results, so as to obtain a reasonable weight
discrete way. Assume that the distance r between the calculation point and thesource term is calculated, the size of each small volume source term after thesource term is discretized is l, and the discrete size of the source term is n is theratio of r to l.
The results show that for the line source, when the discretization size n≥3, thedeviation between the numerical solution and the analytical solution generated
by discretization is less than 1%, which can be ignored. For the volume source,
when the discrete size n≥4, the errors generated by the discretization can bebasically ignored, so the discrete size can be used for dose rate calculation atpoints within this range. In this case, not only can improve the efficiency of calculation, but also can ensure the calculation accuracy.
377

Abstracts - Poster

P4 - 10#4 Environmental Radiation Measurement and Assessment

Distribution of Naturally Occurring Radioactive Materials in
Groundwater of Jeju Island

Sohyeon Lim1, Chung-Hun Han1, Hee-Jung Im1,*
Department of Chemistry, Jeju National University1
Email: imhj@jejunu.ac.kr
Keywords: 222Rn, 40K, Uranium, Thorium, Ground water, Jeju
This study presents distribution of naturally occurring radioactive materials ingroundwater of Jeju island. Radon (222Rn) concentrations were determined onsite by using RAD H2O of RAD7. Potassium (40K) concentrations were performed
by Ion Chromatograph. In addition, the activities of uranium and thorium nuclides
were measured by Inductively Coupled Plasma Mass Spectroscopy. For the analysis of nuclides emitting gamma-rays, High Purity Germanium (HPGe) Gamma
Spectroscopy was used. Groundwater samples were collected from 7 sites ofgroundwater intake facilities for wide area supply in Jaju island from September
to October 2022. All samples are the source of tap water for public water systemsof Jeju island.
The radon concentrations in groundwater were in the range of 0 to 16.4 Bq L-1,
and there was no groundwater exceeding the range of 148 Bq L-1 proposed by
the US EPA. The distribution of uranium and thorium in groundwater varied from0 to 0.5 μg L-1 and 0 to 2.4 ng L-1, respectively. The concentrations of uranium
did not exceed 30 μg L-1, thresholds indicated by the US EPA.
The gamma-emitting nuclides analyzed by HPGe were 210Pb, 212Pb, 214Pb,
40K, 137Cs and 134Cs. The activities of nuclides emitting gamma-rays were
under instrumental detection limit. This result shows that the radiological risks
due to the dose of gamma-emitting nuclides from exposure via groundwater intake is negligible.
By analyzing the concentrations of radon, uranium and thorium, the effective
dose of residents was assessed. The evaluated residents’ effective dose fromradon, uranium and thorium due to intake of drinking water is less than the recommended value of 100 μSv y-1. Consequently, this study indicates that thehealth risks of the residents in Jeju island from naturally occurring radioactive
materials ingested with water is insignificant.

378

P4 - 11

#4 Environmental Radiation Measurement and Assessment#4 Environmental Radiation Measurement and Assessment

Particle Size Distribution of Radioactive Aerosols in Workplace

Feng Zhang1, Pengxiang Li1,*, Ruijun Wang1, Jinhao Li1, Yu Li1, Yongheng Guo1,
Yong Li1
China Institute for Radiation Protection, China1
Email: lipengxiang0214@163.com
Keywords: Workplace, Radioactive Aerosols, Particle Size Distribution, Natural
NuclidesThe concentration and particle size distribution of radioactive aerosol are important parameters in the evaluation of internal irradiation radiation dose. The
particle size distribution of radioactive aerosols produced in various workplaces
is different. It is necessary to establish a research method of particle size distribution of radioactive aerosols produced in workplaces in order to evaluate theinternal irradiation hazards caused by radioactive aerosols scientifically.
Firstly, natural radioactive aerosols in a radioactive laboratory workplace wereselected as the research object. Computational fluid dynamics simulation technology was used to build a model of the selected workplace, and the flow field
simulation was simulated. According to the simulation results, the most representative sampling sites were selected. Then aerosols produced in workplaces
were collected by cascade impactor(Anderson), and the concentrations of uranium and Po-210 in each stage of sampling filter were analyzed.
The results showed that the uranium concentration in the 0-7 stage filtration
membrane ranges from 0.15 to 0.20 ng/m3, without obvious grain size distribution. The Po-210 activity concentration showed a single-peak normal distribution
with the aerosol particle size. The maximum Po-210 activity concentration was
0.11 mBq/m3 on the 7-stage filter with an aerodynamic diameter of 0.43 μm,
accounting for about 37% of the total.
The reason for this phenomenon was that the formation mechanisms of the two
radionuclide aerosols were different. The particulate matter in workplace itself
contained uranium, while the polonium in the aerosol was produced by the adsorption of radon decay daughters(Po-210) on the particulate matter. Polonium
is more easily adsorbed on small particle size aerosol due to its larger specificsurface area. In addition to the formation mechanism of aerosols, the manufacturing process in the workplace also affects the particle size distribution of radioactive aerosols. Future work will focus on the particle size distribution of radioactive aerosols in workplace such as nuclear fuel reprocessing plants and explore the association between particle size distribution and contribution to internal irradiation dose.
379

Abstracts - Poster

P4 - 12#4 Environmental Radiation Measurement and Assessment

Pretreatment Method and Ash Fresh Ratio Calculation of
Biological Sample Analysis in Radiation Environment Monitoring

Jinhao Li1, Feng Zhang1, Ruijun Wang1,*, Qinnan Song1, Yongheng Guo1, Yong
Li1
China Institute for Radiation Protection, China1
Email: huanjingfx@163.com
Keywords: Radiation Environment Monitoring, Biological Sample, Pretreatment,
Ash Fresh Ratio
Radiation environment monitoring is one of the important links to ensure the safe
operation of nuclear facilities. In addition, it is of great significance to evaluate
the quality of regional radiation environment and the dose to the public. It
includes monitoring and evaluating the activity concentration of radionuclides inenvironmental media such as air and surface water and in living organisms in theenvironment surrounding nuclear facilities.
Because of the variety and quantity of biological samples in the environment,
different pretreatment methods should be used for different types of biologicalsamples. In addition, in order to accurately calculate the activity concentration ofradionuclides in biological samples, it is necessary to determine the amount ofsamples collected according to the corresponding ash fresh ratio.
The pretreatment methods of different types of biological samples (fish, shellfish,
livestock and leaf plants) were first introduced, followed by the statistics of thegray-fresh ratio of various biological samples. The results showed that the ash
fresh ratio of fish was in the range of (7.2~27) g/kg and that of shellfish was in
the range of (5.3~18.2) g/kg. The ash fresh ratio of livestock is in the range of(5.3~8.2) g/kg. The ash fresh ratio of leaf plants ranged from (15.9 to 51.2) g/kg,
among which bamboo leaves had the highest ash fresh ratio.
The pretreatment methods should vary depending on the nature of the sample.
For fish, meat, and poultry samples with high oil content, the scales and internalorgans need to be removed first, and then the sample is incinerated andcarbonized. During the above period, attention should be paid to preventing
excessive temperature from causing sample sintering. Due to the different waterand organic matter content, the ash fresh ratio of various biological samples
varies greatly. The statistical results can provide guidance for determining thecollection amount of biological samples in the process of radiation environment
monitoring.

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P4 - 13

#4 Environmental Radiation Measurement and Assessment#4 Environmental Radiation Measurement and Assessment

Cooperation on Radiation Measurements for Cross-Border
Accidents (2) Application of machine learning conversion method
using monitoring data from Fukushima as teaching data

Miyuki Sasaki1, Kotaro Ochi1, Yukihisa Sanada1, Sungyeop Joung2, Yoomi
Choi2, Eunjoong Lee2, Wanook Ji2, Young-Yong Ji2Japan Atomic Energy Agency, Fukushima, Japan1, Korea Atomic Energy Research Institute, Daejeon, Korea2
Email:
Keywords: Fukushima Daiichi Nuclear Power Plant, airborne radiation survey, machine learning
Since the Fukushima Daiichi Nuclear Power Plant accident, some types of radiation monitoring have been conducted in Japan. Walking surveys and unmanned
helicopter surveys have been conducted around the FDNPP to the ambient dose
equivalent rates (air dose rate). The airborne radiation survey (ARS) by unmanned
helicopter has the advantage of measuring large areas including forests. However,
ARS has lower measurement resolution than walking surveys. Sasaki et al. constructed an artificial neural network (ANN) to convert ARS data into air dose rates
at 1 m above ground level using data accumulated after the accident. It has beenreported that the conversion using ANNs can convert to values closer to thosemeasured on the ground than the conventional method. Showing that the ANN
constructed from the Fukushima experience can be applied to detectors otherthan the radiation detector used to construct the ANN will greatly contribute to
the future development of ARS. In this study, we investigated how to apply radiation measurement data acquired with a detector different from the radiation detector used to acquire the training data to an ANN that has already been constructed. For the evaluation, ARS was performed by an unmanned helicopter(FAZER-R-G2, Yamaha Motor Co., Ltd., Shizuoka, Japan) with different detectors. The LaBr3 (Ce) scintillation detector (38 mm φ × 38 mm H × three detectors), which was used to obtain ANN training data, and CZT scintillation detector(15 mm × 15 mm × 7.5 mm), which has different characteristics from LaBr3(Ce),were used for the ARS. The ARS area was 370 m × 270 m, and the average of
air dose rate at 1 m above ground level was 1.45 µSv/h. The measurement altitude was set to about 60 m. The data obtained with the LaBr3 (Ce) scintillation
detector were converted directly in the constructed ANN. The data acquired with381

Abstracts - Poster

P4 - 13CZT have different characteristics from those of LaBr3 (Ce). Therefore, calibration measurements were performed and parameters were created to convert the
CZT count rate to a LaBr3 (Ce) level count rate. The data obtained with CZT wereconverted to LaBr3 (Ce) levels using the created parameters, and the conversion
was performed by ANN.
A comparison of ANN conversion results for LaBr3 (Ce) and CZT measurement
data showed a good correlation (correlation coefficient is about 0.97). This study
shows that the ANN conversion can be applied to data generated with different
detectors by creating conversion parameters.

382

P4 - 14

#4 Environmental Radiation Measurement and Assessment#4 Environmental Radiation Measurement and Assessment

Cooperation on Radiation Measurements for Cross-Border
Accidents (1) A CZT detector-based airborne survey analysis inhigh dose rate environment

Sungyeop Joung1,*, Young-Yong Ji1, Yoomi Choi1, Eunjoong Lee1, Wanook Ji1,
Miyuki Sasaki2, Kotaro Ochi2, Yukihisa Sanada2
Environmental Safety Technology Research Division, Korea Atomic Energy
Research Institute/Korea1, Environmental Monitoring Group, Japan Atomic
Energy Agency/Japan2
Email: jsy1003@kaeri.re.kr
Keywords: Airborne survey, Radiation measurement, CZT detector, Unmanned
aerial vehicle
An airborne survey system named the MARK-A1 was developed to be mounted
on a unmanned aerial vehicle (UAV) for the purpose of application incontaminated areas with high dose rate levels. The MARK-A1 system consists ofa CZT detector, signal processing unit, and positioning and interface units to a
PC on the ground. The weight of the system is below 1 kg for loading on acommercial drone. For experimental verification field testing were conducted inhigh dose rate environment near the Fukushima Daiichi Nuclear Power Plant.
From hovering experiment, the conversion factor was derived for using flat source
mode. Three experiments were conducted to compare the dose rate conversion
results. The first experiment was an airborne survey using a flight with a constant
GPS altitude (flat flight), the second experiment was an airborne survey using aflight with a constant distance between the ground and the UAV as the height of
the ground changes (step-by-step flight). The third experiment was a walksurvey, using a backpack to maintain a constant height of 1 meter from theground. Finally, the dose rate results were compared with those from JAEA
airborne survey analysis data using three LaBr3 (Ce) detector. When comparing
the experimental results, we found that both systems were able to achieve similartendency results as walking, and the results of both systems were similar. In otherwords, the performance of MARK-A1 has been sufficiently verified through thecomparison of the previous experimental results. Therefore, it is expected that
MARK-A1 can be used to quickly scan large and highly contaminated areas thatare difficult for humans to access in the event of a radioactive accident.

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P4 - 14

Acknowledgement
This work was supported by the Nuclear Safety Research Program through the
Korea Foundation Of Nuclear Safety(KoFONS) using the financial resource
granted by the Nuclear Safety and Security Commission(NSSC) of the Republicof Korea. (No. 2205011-0223-CG100)

384

P4 - 15

#4 Environmental Radiation Measurement and Assessment#4 Environmental Radiation Measurement and Assessment

Feasibility Study of Directly Measuring Radon Concentration in Air
Using TPC

Qiuli Zhang1
Department of Engineering Physics, Tsinghua University/China1
Email:
Keywords: radon, time projection chamber, absolute measurement
Radon is a radioactive gas that can cause lung cancer and other health problems
for humans. The concentration of radon in the air can vary depending on thelocation and environmental factors. Therefore, accurate measurement of radon
concentration in air is important for environmental monitoring, radiation
protection, and nuclear physics research. However, most of the current methods
for measuring radon concentration are relative methods with a higher degree of
uncertainty, and the calibration relies on radium standard materials. Time
Projection Chamber(TPC) has the ability of high-precision three-dimensional
track detection, and can provide the momentum and energy loss information ofparticles for particle identification.
In this paper, we establish a simulation process for directly measuring the radon
concentration in air using TPC based on the Monte Carlo simulation software
Geant4 and Garfield++. Geant4 is mainly used to simulate the decay of radon,
the transport and ionization of alpha particles in gas, while Garfield++ is used tosimulate the drift and avalanche of electrons, and the final signal readout. To
ensure direct measurement of the radon concentration without any intermediate
processes, the working gas of the TPC must be clean air. The simulation also
involved optimizing the TPC detector parameters for radon measurement,
including the electric field, the detector geometry, and the readout system. By
combining the energy and trajectory information, the number of alpha particles
can be obtained, thus removing environmental background such as gamma rays
and cosmic rays to improve the detection sensitivity.
Based on the simulation results, the feasibility analysis of measuring radon
concentration in air directly using TPC was presented. Compared with traditional
methods of measuring radon concentration, the method proposed in this paper
has advantages such as high sensitivity, low background, and easy calibration.
It provides a new idea for absolute measurement of radon concentration in air
and contributes to the development of more accurate and reliable methods forradon measurement.

385

Abstracts - Poster

P4 - 16#4 Environmental Radiation Measurement and Assessment

Life cycle greenhouse gas emissions from nuclear
power chain in China

Feifei Wu1, Yan Wang1, Jing KANG1, Bing Lian1,*
China Institute for Radiation Protection, Key Laboratory of Radiation
Environment & Health of the Ministry of Ecology and Environment, Taiyuan,
030006, China1
Email: lianbing00@ sina. com
Keywords: Process-based life cycle assessment, nuclear power, green gas
Nuclear power, regarded as a kind of green energy, plays an important role incarbon emission reduction. A unified and standardized greenhouse gas (GHG)
emissions accounting framework for the nuclear power plant has not been established. Different views existed on the accounting boundaries, GHG types, accounting methods and other aspects in existing studies, which was not conducive to the comparative study of accounting results.
The GHG framework of nuclear power was established based on the life cycle
theory . In this framework, both space and time boundaries were considered in
the system boundary, and GHG types included carbon dioxide (CO2), methane(CH4) and nitrous oxide (N2O). The process-based life cycle assessment methods were used to calculate the GHG emission of a pressurised reactor in China.
Based on the phase division of nuclear power life cycle, GHG with different types
were calculated. The escaped CH4 should be considered during uranium mining,
thus the CO2 equivalent was calculated with CH4 and CO2 emission. During
Uranium milling, conversion and enrichment as well as decommissioning phases,
CH4 and N2O were released except for CO2. Moreover, CO2, CH4 and N2Oemissions shall be calculated in the waste disposal phase.
The results indicated that the GHG emission of non-carbon dioxide (CH4 andN2O) accounted for about 17% of the total greenhouse gas emission using the
prepossess-based life cycle assessment. Using the process-based method, thegreenhouse gas emission ranged between 8.56~10.69 gCO2 e/kWh. In addition
to system boundary, greenhouse gas types and assessment method, the operating life, technology and energy consumption level of nuclear power plants weremajor factors that affected the accounting results of greenhouse gas emissions.
The results have certain reference values for the policy and strategy formulations
for facilitating the development of nuclear power generation.

386

P4 - 17

#4 Environmental Radiation Measurement and Assessment#4 Environmental Radiation Measurement and Assessment

Study on Optimal Control of Liquid Effluent Emission from Small
Nuclear Power Reactor

Meng Wang1, Bing Lian1,*, Yan Wang1, Jing Kang1
China Institute for Radiation Protection,Key Laboratory of Radiation Environment& Health of the Ministry of Ecology and Environment, Taiyuan, 030006，China1
Email: lianbing00@sina.com
Keywords: Small nuclear power reactors，Radioactive liquid effluent discharge，Optimize
Small nuclear power reactors have the advantages of low investment, short construction period, high site selection flexibility and good safety performance, andvarious countries are accelerating their application. At present, China has notestablished an environmental impact assessment standard system for cryogenic
heating reactors, and the emission limit of radioactive liquid effluent from small
nuclear power reactors has not yet been stipulated. In this paper, taking the
2×200MWt HAPPY200 small reactor in City A as an example, the multi-attribute
effect analysis method and the fastest decline method were used to optimize thedischarge method of radioactive liquid effluent into the receiving water. The optimal control of radioactive liquid effluent discharge from small reactors needs tocomprehensively consider environmental, economic and social factors, and thispaper comprehensively considers the influence of factors such as individual effective dose of key resident groups, individual effective dose of occupational
personnel, public collective effective dose, waste treatment cost and public response, and obtains the willingness index of different stakeholders and relatedparties to consider the control of radioactive liquid effluent discharge from small
reactors through questionnaire survey. Using engineering judgment and multiattribute effect function analysis methods, the utility function value of radioactive
liquid effluent discharge under different influencing factor weight values is calculated, and the relationship curve between radioactive liquid effluent discharge
and utility function value is obtained by least squares fitting. Finally, the multiattribute effect function analysis method and the fastest descent method are effectively combined to obtain the optimal value of radioactive liquid effluent emission from small reactors. This research method is innovative and breakthrough inthe field of environmental radiation protection, and the optimization results obtained are of great significance for the establishment of the environmental assessment standard system of small nuclear power reactors and the determinationof the emission limit of radioactive liquid effluent, and can be used as an important basis for the decision-making of the audit authority.

387

Abstracts - Poster

P4 - 18#4 Environmental Radiation Measurement and Assessment

Determination of 226Ra in drinking water in Southern China by
Liquid Scintillation Counting
CHEN GUO, HAILAN YANG*, YOUKUN YANG, LI BAO, RUIJUN WANG
China Institute for Radiation Protection1
Email: 10338728@qq.com
Keywords: LSC,radium-226,drinking water
The determination of Ra-isotopes in environmental samples is very importantfrom the radiological point of view. Due to their radiotoxicity, especially that of
226Ra, the contamination hazard is potentially dangerous to human beings evenat low concentration levels. For the general public, the consumption of food andwater containing radium may lead to the accumulation of these nuclides, contributing to radiological dose.National and international regulators have shown an
increasing interest in regulating radium levels in most industries and consumer
products. The US Environmental Protection Agency (EPA) has released national
drinking water regulations which limit the concentration of various contaminants
to defined levels and set the maximum contaminant level for the sum of 226Ra
and 228Ra to 185 mBq L-1 (5 pCi L-1). European Commission Drinking Water
Directive 98/83/EC gives an indicative effective dose of 0.1 mSv a-1 from ingestion of radioactive substances in tap water. Therefore, monitoring of 226Ra
in drinking water is essential.
Radium-226 was measured in drinking water in southern China by LSC. Evaporing water samples to concentrate, then mixed with a immiscible scintillation
cocktail which can separating 222Rn from 226Ra and 210Pb in the aqueous solution. When pre-concentration is desired, acidify the sample to between pH 1
and pH 3 with HNO3 After a period of time, when 226Ra has reached equilibriumwith its substrates, alpha/beta discrimination liquid scintillation counting are performed to separate 222Rn and its alpha substrates 218Po and 214Po from thebeta substrates 214Bi and 210Pb counts, and the activity concentration of 226Ra
in the aqueous phase is indirectly calculated by measuring the total alpha countsof 222Rn, 218Po and 214Po.
Nine sites of drinking water were sampled for determination of 226Ra, and theresults indicated that the content of 226Ra in drinking water in southern China is
betweenIn this work, a method is given for the measurement of 226Ra in waters by liquid
scintillation counting. The activity concentration of 226Ra in the aqueous phaseis indirectly calculated by measuring the total alpha counts of 222Rn, 218Po and214Po by separating 222Rn from the aqueous solution with a immiscible scintillation cocktail. No special radiological significance of the studied waters has
been found.
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#4 Environmental Radiation Measurement and Assessment#4 Environmental Radiation Measurement and Assessment

Combined transuranic-nuclide analysis.of aerosol samplese

Ke Du1, Peng Xiang Li1, Xu Yuan Ma, Yu Hu Han, Feng Zhang, Li Bao, Rui Jun
Wang*
China Institute for Radiation Protection1
Email: huanjingfx@163.com
Keywords: Aerosol, Transuranium nuclide,Resin, Joint analysis.
Fuel cladding breakage in PWR nuclear power plantscan result in fuel releases
from fuel assemblies into the coolant, where α radionuclides, dominated bytransuranic elements, can cause a contamination of systems, equipment, andjob sites. Among them, 238~241Pu, 237Np, 241Am, 244Cm are long-lived
extremely toxic α transuranic nuclides, and are one of the key and difficult points
in monitoring and protection of radioactive aerosols in the operation site. Due tothe similar energy of 238~241Pu, 237Np, 241Am, 244Cm, and the extremely low
content of transuranium nuclides in the gas soluble gel, sampling is difficult, andthe existing analytical methods are not sensitive enough to meet the requirementsof quality control timeliness.
Ion exchange method is one of the commonly used separation and purification
methods for nuclide isotope analysis, with advantages of easy operation and low
dependence on analysts and equipment. Moreover, the structural stability andreusability of the ion exchange resin itself are also very good. In order toeffectively improve the separation and purification effect of nuclides, avoid
interference from other nuclides during the measurement process, and reduce
the separation time.Based on the different adsorption effects of nuclides withdifferent valence states on the resin, the combined analysis of 238~241Pu,
237Np, 241Am, 244Cm was achieved by using TEVA resin, UTEVA resin, and
DGA resin.
The analysis results showed that the recoveries of238~241Pu, 237Np, 241Am,
244Cm in aerosol samples were≥80%, and the separation and purification time
was less than 8h. Compared with the traditional method, this method effectively
saves manpower and material resources, reduces the harm of radioactivity tooperators, and shortens the working time.
In the joint analysis of transuranium nuclides in aerosol samples, a new method
for separating neptunium, plutonium, americium, and curium from the same
sample has been established, which has a positive promoting effect on theseparation and determination of other radioactive solutions.

389

Abstracts - Poster

P4 - 20#4 Environmental Radiation Measurement and Assessment

The Content of Tritium and Carbon-14 in Environmental Medium
around Shidaowan High Temperature Gas Cooled Reactor NPPQin Nan SONG1,*
China Institute for Radiation Protection/China1
Email: 649111459@qq.com
Keywords: Shidaowan HTGCR NPP，Tritium，Canbon-14，Environmental medium
Shidaowan Nuclear Power Plant is the world's first fourth-generation high temperature gas cooled reactor nuclear power plant put into commercial operation.
The main source nuclides in its operation period are tritium and carbon-14. It is
uncertain whether the nuclear power plant has an impact on the surrounding
environment after operation.
In order to judge the impact of Shidaowan Nuclear Power Plant on the surrounding environment after operation, tritium and carbon-14 in the environmental medium within 10km after operation were monitored and analyzed for a whole year,
and ambient air, seawater and marine organisms were collected in four quarters.
The average content of HTO in air is 0.58 Bq/L, the average content of C-14 in
air is 0.23 Bq/gC, the average content of HTO in seawater is 0.31 Bq/L, and theaverage content of C-14 in seawater is 0.19 Bq/gC. There is no abnormality inthe measurement results of 9 marine organisms such as mackerel, oyster andkelp.
The monitoring results show that, compared with the background investigation
results, tritium and carbon-14 in the environmental medium have not increased
significantly after the operation of the nuclear power plant, and have no obvious
impact on the surrounding environment. The impact of radionuclide emissions
from the nuclear power plant on the surrounding environment is controllable.

390

P4 - 21

#4 Environmental Radiation Measurement and Assessment#4 Environmental Radiation Measurement and Assessment

Assessment of Uncertainty in Collecting Radioactive Soil Samples
by Sampling and Analytical Quality Control（SAX）

Ruijun Wang*, Qinnan Song, Li Bao, Jinhao Li, Xvjia Hou, Haiyang Wu, Qiqi Shi
Key Laboratory of Radiation Environment & Health of the Ministry of Ecology
and Environment, China Institute for radiation proctection /China1
Email: wang.ruijun@163.com
Keywords: SAX Method, Sampling, Uncertainty
Sampling uncertainty is an important part of the uncertainty in radiation
environmental monitoring results, Because of the difficulty of quantify sampling
uncertainty, at present, the uncertainty of monitoring results of radiation
environment only includes the analysis process, without considering the sampling
process. Sampling design using SAX method and data analysis using robust
variance analysis method can quantify the uncentainty of sampling process. A
batch of actual monitoring data are analyzed by the SAX method, and obtained
the sampling uncertainty. And studied the uncertainty carefully,.It is consideredthat SAX method can be used to guide the monitoring of radiation environment,so as to optimize the monitoring strategy and improve the monitoring quality.
Variance is a measure of the degree of dispersion of a group of data. It is
generally believed that the degree of dispersion measured in a batch of samples
does not only come from the analysis process, but comes from the following
three aspects: (1) The inherent variance caused by the different "true values"
measured in different samples is commonly referred to as geochemical variancein geochemical research, and in this article, it is referred to as inherent variance,
which is represented by ; (2) For the same sampling point, during actual sampling,different specific sampling locations or sampling strategies lead to different
measured "true values" in different samples collected. The variance generated bythis factor occurs during the sampling process, so it is called sampling variance,
which is represented by ; (3) The variance generated by the sample during theanalysis process is called the analysis variance, which is represented by .
In a radiation environment monitoring, L sampling points are arranged for the soil
medium within the monitoring range, and m parallel samples are collected at eachsampling point. After each sample enters the laboratory, n parallel analyses are
performed. The difference between n parallel analysis results for each sample
can be considered to be due to analytical error; The difference between the
results of m samples at each point is considered to come from sampling error

391

Abstracts - Poster

P4 - 21and analysis error, while the difference between the results of points and points
is considered to come from inherent inherent error, sampling error, and analysis
error; Therefore, theoretically, analysis errors can be solved from multiple parallel
analysis results, and sampling errors and analysis errors can be solved frommultiple parallel sample results, thereby obtaining sampling errors; Similarly, theinherent error, sampling error, and analysis error can be solved from the resultsof multiple points, and the inherent error can be obtained by using the previously
obtained analysis error and sampling error. Finally, the values of analysis error,
sampling error, and intrinsic error can be obtained.

392

P4 - 22

#4 Environmental Radiation Measurement and Assessment#4 Environmental Radiation Measurement and Assessment

Evaporation behavior of polonium-210 from lead-bismuth
eutectic at low temperature

Huiyun Xue1, Yong Guo2, Ping Xu1, Guobing Yu3, Yujia Kang1, Zhi Chen1,*
University of Science and Technology of China, Anhui Province, China1, China
Nuclear Power Technology Research Institute, 518000 Guangdong, China2,
Anhui Radiation Environmental Supervision Station, Anhui Province, China3
Email: zchen@ustc.edu.cn
Keywords: polonium-210, evaporation
Lead-bismuth eutectic (LBE) has been utilized as a spallation target or coolant
in the nuclear industry, and the accumulation of polonium-210 in LBE, producedby neutron-activated bismuth, presents a critical issue in safety operation. It isnecessary to evaluate the release of polonium-210 in LBE for safety evaluations
during operation, maintenance and accidental conditions. The transpiration
method, a dynamic process of measuring the activity of condensed vapor or
sample, is frequently used to characterize polonium-210 evaporation from LBE.
In the past few decades, researchers have found that the Henry’s law constantsof polonium-210 evaporation at high temperatures are consistent with the
Antoine-type correlation. However, deviations from that evaporation regime werediscovered at low temperatures. The parameters promoting the enhancement ofpolonium-210 volatilization should be systematically studied in the subsequent
period.
In this study, the activity and homogeneity of polonium-210 in irradiated LBE
were determined. FISPACT-2007 was used to simulate the activation of 210Po
in neutron-irradiated LBE, and the neutron-irradiated LBE was determined byalpha spectrometry based on the ISO 13161:2020 method. After that, a set ofexperimental results of polonium-210 evaporated in a temperature range of 100600℃ were measured to have a better fundamental understanding of the potential
mechanism. The evaporation from LBE was measured through the transpiration
method.
Experimental results of the activities of polonium-210 produced in neutronirradiated LBE show that the measured values of the three batches of samples
were generally consistent with the simulation values. At the same time, thenonuniform distribution of activated polonium-210 in LBE was observed in view393

Abstracts - Poster

P4 - 22of its volatility and migration characteristics. Considering the nonuniform
distribution of polonium-210, the experimental results of polonium-210
evaporation show that the Henry constants of polonium-210 in this temperature
range were several orders of magnitude higher than the expected values fromearlier studies at higher temperatures. Based on these data, the thermochemical
properties of polonium-210 are further estimated to explain the phenomenon ofincreased evaporation at low temperatures, which offers an observation ofpolonium behaviors during the operation period of LBE-based facilities.

394

P4 - 23

#4 Environmental Radiation Measurement and Assessment#4 Environmental Radiation Measurement and Assessment

Wind tunnel study of effects of thermal stratification on turbulent
flow and diffusion over a three-dimensional hill

Yun-peng LI1, Ruo-jie LI1, Yan-hui Pan1, Si-bo Lyu1, Zhan-jie Guo1,
Jun-fang Zhang1
China Institute for Radiation Protection1
Email:
Keywords: Wind tunnel, thermal stratification, turbulent flow, diffusion, three-dimensional hill
In the environmental impact assessment of nuclear power plants, it is necessary
to quantitatively give the impact of airborne radionuclides on the surrounding
environment and the public under normal operating conditions and accident conditions of nuclear power plants. The dispersion characteristics of airborne pollutants in the vicinity of nuclear power plants are of great significance to the site
selection, construction, operation and decommissioning of nuclear power plants.
The influence of terrain complexity on the structure of atmospheric boundary
layer depends largely on the local thermal stratification. Thermal stratification affect the height and structure of the atmospheric boundary layer as well as thevelocity, temperature and turbulence intensity profiles in the boundary layer,
which play a major role in pollutant dispersion at the actual atmosphere. Therefore, it is of great concern to evaluate the impact of pollutant diffusion on theenvironment and the public on complex terrain under different thermal stratifications.
This experiment is carried out in the atmospheric boundary layer wind tunnel of
the China Institute for Radiation Protection, which is equipped with a thermel
stratification simulation system. The simulation of the non-neutral atmosphericboundary layer can be realized by adjusting the inlet temperature profile. Windtunnel experiments are used to study the turbulent structure of the flow and pollutant diffusion around the three-dimensional hill in the atmospheric boundary
layer. The influence of neutral and stable atmospheric boundary layer (neutral:Reδ＝2.15×104,Riδ＝0; stable: Reδ＝2.0×104,Riδ＝0.76)on flow and diffusion
is investigated. Velocity, turbulence intensity and temperature is measured by
combining cold wire and hot wire. CO is used as tracer in concentration field.
The results show that：under the neutral and stable layer stratification, the threedimensional hill cause a significant decrease of the near-surface velocity andturbulence intensity in the wake region. When atmospheric stratification is stable,

395

Abstracts - Poster

P4 - 23

the longitudinal velocity and turbulence intensity is lower than under neutral conditions. The pollutants appear to wash down in the vertical direction. The stable
thermal stratification inhibits the diffusion of pollutants in the vertical direction,
which causes an increase in concentrations. The maximum normalized concentration appears near the three-dimensional hill. And compared with neutral thermal stratification, the height of the concentration peak is lower.
Acknowledgement
This research was funded by the National Natural Science Foundation ofChina( U21A20524) and supported by the special fund for Science and Technology Innovation Teams of Shanxi Province(202204051002026).

396

P4 - 24

#4 Environmental Radiation Measurement and Assessment#4 Environmental Radiation Measurement and Assessment

Background Spectrum Simulation Development and Verification
for In-Situ Measurement Technology Application

Byungchae Lee1,*, Jihyun Yu1, Jonghoa Kim1, Jangsoo Suh1, Sangbum Hong2
R&D Institute, Sae-An Enertech Corp/Korea1, Korea Atomic Energy ResearchInstitute/Korea2
Email: bclee@sae-tec.com
Keywords: Background spectrum simulation, Radioactivity depth distribution,
Peak to compton method
In order to reduce the measurement uncertainty of In-situ measurement technology, the influence of the environment must be minimized. In general, shields
are used to minimize environmental impacts. However, there are difficulties infield application due to the weight of the shield. Therefore, it is necessary todevelop In-situ measurement technology suitable for field application[1, 2].
In this research, the background spectrum was simulated to reduce measurement
uncertainty of the Peak to Compton(PTC) method, which is an In-situ measurement technology developed in previous research[3, 4]. In the background spectrum simulation, 34 gamma ray energies emitted from natural nuclides were simulated by MCNP. The radioactivity depth distribution was calculated by applying
the simulated background spectrum and the PTC method to three sites in Jeju.
The accuracy of the PTC method was evaluated by comparing the results ofanalyzing the radioactivity depth distribution by collecting samples from the same
point.
As a result of comparison with the sample analysis results by applying the background spectrum simulation and the PTC method, the average relative error of
the three measurement points was 4.00 %. As a result of comparing with thesample analysis results by applying only the PTC method, the average relative
error of the three measurement points was 13.34 %. In the case of applying thebackground spectrum simulation, the relative error decreased by 9.34 % compared to the case of not applying the background spectrum simulation. Therefore,
it is judged that measurement uncertainty can be reduced if the background
spectrum simulation method is applied together with the PTC method.
In order to reduce the measurement uncertainty of the Peak to Compton method,an In-situ measurement technique, background spectra were simulated with
MCNP. By applying the background spectrum simulation method, the radioactivity
depth distribution can be analyzed within 4% relative error. In order to apply the397

Abstracts - Poster

P4 - 24

evaluation method of this study to the dismantling site, verification in various
environments is required.
Reference[1] S.B. Hong, J.S. Nam, Y.S. Choi, B.K. Seo, J.K. Moon, Application of In Situ
Measurement for Site Remediation and Final Status Survey of Decommissioning
KRR Site, Journal of Radiation Protection and Research 41(2), 2016. [2] International Commissioning on Radiation Units and Measurement (ICRU), ICRU Report No. 53, 1994. [3] B.C. Lee, S.B. Hong, B.K. Seo, J.H. Kim, Y.U. Kim, In situ
measurement of Cs distribution in the soil, New Physics. 69(7), pp. 701-706,
2019. [4] B.C. Lee, Y.U. Kim, W.S. L’yi, J.H. Kim, B.K. Seo, S.B. Hong, Radiological analysis for radioactivity depth distribution in activated concrete using
gamma-ray spectrometry, Applied Radiation and Isotopes 169, 2021.

398

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#4 Environmental Radiation Measurement and Assessment#4 Environmental Radiation Measurement and Assessment

Sensitivity Analysis of Distribution Coefficient by Nuclide
According to Landfill of NORM Waste

Seong Yeon LEE1, Seong Hun JEON1, Min Seong KIM1, Kwang Pyo KIM1,*
Department of Nuclear Engineering, Kyung Hee University/Korea1
Email: kpkim@khu.ac.kr
Keywords: NORM, Distribution Coefficient, RESRAD, NDD
In 2018, social issues arose in Korea due to bed containing monazite, and public
interest in NORM emerged. The Nuclear Safety and Security Commission
announced that it would collect and dispose of all radon beds, and the Ministryof Environment prepared a legal basis for NORM waste disposal by revising the
Waste Management Act. When NORM waste is disposed, there is a possibility of
causing radiological impact through various pathways such as direct externalradiation and groundwater leaching. Therefore, the radiological impacts of NORM
waste landfill are determined by various factors, and accordingly, in order toperform a reliable assessment, it is necessary to select and reflect the main
exposure factors in the assessment. Therefore, in this study, a sensitivity analysis
on the distribution coefficient (Kd) of natural radionuclides was performed as apreliminary study for the development of major exposure factors for theradiological impacts of NORM waste landfill. In this study, the sensitivity analysisof the Kd for natural radionuclides was performed using the RESRAD-ONSITE.
Sensitivity analysis was performed in the order of (1) selection of nuclides, (2)selection of exposure scenarios, (3) analysis of sensitivity of Kd by nuclide. In
selection of nuclides, U-238 series, Th-232 series, K-40 which are generally
contained in NORM waste were selected. As the exposure scenario, the mostconservative scenario, the resident farmer scenario was selected. Sensitivity
analysis used NDD methodology, and factors with NDD values exceeding 50 were
derived as sensitive factors. The NDD values of the Kd of U-238, U-234, andTh-230 were derived as 73.5, 64.9, and 55.7, indicating that they were sensitive.
And as a result of analyzing the change in dose by exposure path according to
the change in the Kd for sensitive nuclides, it was found that the internal radiationdose due to inhalation, intake of meat, milk increased as the Kd increased. This
indicates that the larger the Kd, the more radionuclides exist in the contaminatedzone, and that U-238, U-234, Th-230 derived as sensitive nuclides increase
their contribution to the internal radiation dose because they are alpha-emitting
nuclides. In addition, the NDD value of the Kd for Th-228, Pb-210, Ra-228 was399

Abstracts - Poster

P4 - 25

derived as less than 3, indicating that it was not sensitive and that the impact on
the radiation dose was insignificant. In this study, a sensitivity analysis was
performed on the Kd of natural radionuclide according to NORM waste landfill.
For the sensitivity analysis, the RESRAD-ONSITE was used, and the sensitivity
analysis for the Kd was performed using the NDD methodology. As a result of thesensitivity analysis, the partition coefficients for U-238, U-234, and Th-230
nuclides were derived to be sensitive, and it was found that the internal radiationdose such as inhalation and intake increased as the Kd increased. The results ofthis study are expected to be used as basic data when assessing the safety ofradiological Impacts that may occur during landfill of NORM waste in Korea.
Acknowledgement*This work was supported through the KoFONS using the financial resource
granted by NSSC. (No. 2105035)

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P4 - 26

#4 Environmental Radiation Measurement and Assessment#4 Environmental Radiation Measurement and Assessment

A Study on Radiological Characterization for Alpha Source Terms
in Decommissioning Nuclear Power Plants

Seong Jun Kim1, Tae Young Kong 1,*
Department of Nuclear Engineering, School of Chosun University/Korea1
Email: tykong@chosun.ac.kr
Keywords: Radiological Characterization, Decommissioning, alpha
Kori Unit 1, the first commercial Korean nuclear power plant (NPP) permanently
shut down in 2017, submitted a letter of approval for decommissioning to the
Nuclear Safety and Security Commission (NSSC), the regulatory body, in May
2021. The decommissioning approval includes information on the amount andtype of radionuclides present in the NPP, its distribution, and the physical andchemical conditions using radiological characterization (RC). The RC results
enable the calculation of the cost of decommissioning and the establishment of
the decommissioning plan. The RC results also can be used as data for radiation
protection to keep occupational dose as low as reasonably achievable.
For the safe decommissioning of Kori Unit 1, the first thing to consider isminimizing the exposure dose to workers. In decommissioning NPPs, structures,
systems, and components (SSCs) are cut and dismantled. When dismantling theactivated SSCs, substances are suspended in the air, which may cause internal
exposure to workers. These radioactive aerosols contain gamma, beta, and alphanuclides. If the workers inhale the alpha particles during the decommissioning
works, he or she may have serious internal exposure due to the highest radiation
weighting factor and long half-lives. This study analyzed the RC process of thedecommissioning NPPs and proposed a characterization methodology for alphasource terms to minimize the internal exposure of workers.
Acknowledgement
This research was supported by the Nuclear Safety Research Program throughthe Korea Foundation of Nuclear Safety using the financial resource granted by
the Nuclear Safety and Security Commission of the Republic of Korea (No.
2106024-0323-SB110).

401

Abstracts - Poster

P4 - 27#4 Environmental Radiation Measurement and Assessment

Deep learning-based spectrum-dose conversion model for KAERIMARK-M detection system
Eunjoong Lee1,*, Byoungil Jeon1, Young-Yong Ji1
Environmental Safety Technology Research Division, Korea Atomic Energy
Research Institute, Korea1
Email: leej0715@kaeri.re.kr
Keywords: Spectrum-dose conversion, Deep learning, Environmental radiation
monitoring, Monte-Carlo simulation
After the Fukushima nuclear power plant accident, KAERI(Korea Atomic Energy
Research Institute) and JAEA(Japan Atomic Energy Agency) conducted several
joint surveys of environmental radiation monitoring. The ambient dose rate overa wide area was evaluated and the dose rate distribution map was obtained. The
dose rate was evaluated using G-factor (G(E) function) during the survey. Thismethod converts the spectrum to dose rate in real-time with high accuracy.
KAERI has developed the MARK series(Monitoring of Ambient Radiation of KAERI)
for radiation monitoring and the G-factor for each system was derived. The MARK
system has high dose rate evaluation accuracy. However, during the airborne or carborne survey, the uncertainty of the measured spectrum increases since the detector
moves rapidly and short detection time. Accordingly, the dose rate can differ fromthe actual value. In addition, in the G-factor calculation process, errors may occur
considering the angles of incident radiation. Therefore, to solve these problems, a
deep learning-based spectrum to dose conversion model was developed.
The MARK-M(Multipurpose survey system, 2-inch LaBr3(Ce)) model was selected for the detection system. To train the deep learning model, ambient dose
equivalent(H*(10)) and gamma spectrum were calculated by MCNP(Monte Carlo
N Particle) code. The detection system was modeled inside as a spherical source
to consider the effect of incident angle. During the monitoring, detection uncertainty exists in measurement because the data acquisition period is as short asseveral seconds. Accordingly, spectrum data containing noise was generated.
10,000,000 data pairs with arbitrary energies were generated and used as data
set for the deep learning model. As a result, the deep learning model successfully
predicted the ambient dose rate from spectra with over 95% accuracy. Based on
experimental data at Fukushima, the dose assessment results of JAEA, the existing system of KAERI, and deep learning models will be compared.
Acknowledgement
This work was supported by a grant from the National Research Foundation of
Korea(NRF), funded by the Korean government, Ministry of Science and ICT(RS2022-00144210).
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P4 - 28

#4 Environmental Radiation Measurement and Assessment#4 Environmental Radiation Measurement and Assessment

Study of the radiation effects of radioactive residues in soil at a
decommissioning site

ZhiXiang Yu
China Institute for Radiation Protection1
Email:
Keywords: radiation impact assessment; RESRAD procedure; decommissioning
sites
Decommissioning site contamination is highly insidious, hysteresis and persistent,
and site contaminants are usually present in and transferred through the soil,
which change and move very slowly, and the receptors only become potentially
radiologically hazardous when they are exposed to contaminants for a long time.
Therefore, in this paper, we analyze the distribution of site contaminants, thedegree of contamination and the main contaminant (uranium), establish aconceptual model of the contaminated site, use the RESRAD program tocalculate the acceptable activity concentration of the main contaminant (uranium)
remaining in the soil of a decommissioned site, and divide the residents into age
groups to further study the radiation dose of contaminants to residents ofdifferent age groups, and finally arrive at the radiation impact on different age
groups. The radiation effects of the different age groups were finally derived.

403

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P4 - 29#4 Environmental Radiation Measurement and Assessment

Analysis of Carbon-14 Radioactive Effluents Released from
Korean Nuclear Power Plants and Its Resulting Public DoseHwa Pyoung KIM1,*, Tae Young KONG1
Chosun university Department of Nuclear Engineering / Korea1
Email: khp0601@chosun.kr
Keywords: PWR, PHWR, Carbon-14, the public dose, radioactive effluents
Currently, there are 24 nuclear power plants (NPPs) in Korea, including 21 pressurized water reactors (PWRs) and 3 pressurized heavy water reactors (PHWRs).
Korean NPPs release radioactive effluents into the environment during the operation. Radioactive effluents released from NPPs contain various radioactive materials, such as 3H, 14C, noble gas, particulates, and iodine.
Of the radioactive effluents released from Korean NPPs, 14C is a pure beta emitter with a half-life of 5,730 years. In addition, 14C is a potential nuclide that canbe inhaled and ingested by the human body to cause internal exposure. Currently,
the amount of 14C present in the environment is gradually decreasing, but therelease of 14C from NPPs is increasing. 14C released to the environment can beinhaled and ingested by the public around NPPs through various pathways. It isvery important to monitor and manage 14C released into the environment because it contributes the dose of public around NPPs.
It is inevitable to release radioactive effluents while operating NPPs. Korea Hydro& Nuclear Power manages and monitors radioactive effluents. This study compared the 10-year average of 14C released from Korean NPPs depending on thereactor types, such as PWRs and PHWRs, and analyzed the 10-year (2012-2021)
average of the public dose due to 14C released from Korean NPPs.
As a result, 14C released from PWRs accounted for 3.1% of the total release from
PWRs during the recent 10 years (2012-2021). For PHWRs, 14C release accounted for 1.12% of the total release. In addition, the 10-year (2012-2021)
average of public dose due to 14C released from Korean NPPs showed that 14C
accounted for more than 90% and 80% of the total public dose at PWRs and
PHWRs, respectively. The results of public dose due to 14C from 2012 to 2021
show that the public dose was gradually increasing every year. Based on theseresults, 14C monitoring and management are important in the future to reducethe 14C release and its resulting public dose.

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#4 Environmental Radiation Measurement and Assessment#4 Environmental Radiation Measurement and Assessment

Specification for monitoring of uranium aerosol concentration in
the air at workplaces of uranium processing and fuel fabrication
facilities

Baoli WU1,*, Xiaona GU1, Xiangming XUE1, Kai YANG1
China Institute for Radiation Protection/ China1
Email: wubaoli@126.com
Keywords: uranium,aerosol,uranium processing and fuel fabrication
facilities,workplace
With the continuous development of the nuclear industry, the occupational
hazards of nuclear facilities to the staff have attracted more and more attention.
As an important nuclide in nuclear industry, uranium is involved in almost all
aspects, such as uranium mining, ore processing (hydrometallurgy), uranium
refining, enriched uranium production and fuel element manufacturing. Nuclear
workers should pay attention not only to the radiation hazards of uranium, butalso to the chemical hazards of uranium in the case of acute ingestion of solubleuranium compounds. According to the radiation protection regulations foruranium processing and fuel fabrication facilities (EJ 1056-2018), the workplaceshould routinely monitor the concentration of uranium aerosols in the air.
In this paper, the pre-sampling field investigation, representative sampling point
selection, sampler preparation, uranium aerosol sample collection, sample
pretreatment and sample detection condition optimization in the workplace ofuranium processing and fuel fabrication facilities are studied, the standard formonitoring the concentration of uranium aerosols in the air of uranium processing
and fuel fabrication facilities was established.
The specification for monitoring the concentration of uranium aerosol in the airof workplace is divided into two steps: sample collection and sample laboratory
analysis. In the first step, field investigation was carried out on uraniumcontaining raw materials, auxiliary materials, input mode of raw materials,
production process, working mode of workers, frequency and duration of
exposure to uranium aerosols, diffusion law and existing state of uranium
aerosols, etc. , and according to the survey content to select a representative
sampling duty station. Select the appropriate sampler at the sampling site forsample collection. In the second step, the uranium aerosol samples wereanalyzed in the laboratory. The filter membrane was ashed and dissolved with405

Abstracts - Poster

P4 - 30

nitric acid, the pH of the sample solution was adjusted, and the sample wastested by laser-fluorescence method. The detection limit, precision and recoveryof the method were tested. The relative standard deviation (RSD) was less than
5%, and the recovery was 96-104%.
The method is easy to operate and the analysis result is stable and reliable. The
method has been applied to several typical uranium processing and fuelfabrication facilities with satisfactory results.

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#4 Environmental Radiation Measurement and Assessment#4 Environmental Radiation Measurement and Assessment

Review of Dose Estimation Considering Decay Correction for theDecommissioning of Nuclear Power Plants

Hyung-Woo Seo*, Ji-Hwan Yu, Gi-Lim Kim
Decommissioning Technology Group/KHNP CRI1
Email: sehywo1727@khnp.co.kr
Keywords: Decommissioning, Dose estimation, Decay correction
In the decommissioning of Nuclear Power Plants (NPPs), the estimation ofradiation exposure dose due to the activities of workers is an essential task to be
performed in establishing a decommissioning plan. Typically, in the case of alicensee with experience in decommissioning, it will be possible to calculate theexpected exposure dose for each activity based on prior experiences. However,if there is no prior experience, it can be measured to some extent throughcomparison with overseas cases. Therefore, this study analyzes the decay effectbased on the experience data of overseas cases and predicts the results
according to the assumption of dose prediction of domestic decommissioning
NPP.
For the dose estimation of each NPP, one of the important factors is to consider
the effect of the half-life of radionuclides. The attenuation due to the decay ofradionuclides can be expressed by the decay constant that can be expressedagain as a function of half-life (T). In this study, the decay period was normalized
to compare the dose results between decommissioned NPPs.
Changes in decay correction coefficient were calculated for three radionuclides,
and it was found that the decay correction coefficient value increased as the
decay correction year decreased to the negative direction. In addition, the shorter
the half-life of the representative nuclide, the larger the decay correction
coefficient even if the value of the decay correction year is the same. In case of
Kori Unit 1, normalized dose was calculated for the assumed initial dose if thehigh radiation dose work was expected to be performed around 2030.
In order to calculate the normalized dose in the evaluation of worker dose thatmust be performed to establish a plan for dismantling the NPP, this study
examined the effect of the decay correction year. By comparing the start time ofthe high radiation dose work from the base NPP, it was possible to observe the407

Abstracts - Poster

P4 - 31change in the normalized dose accordingly. As a result, preliminary doses werereviewed in consideration of the decay correction year to compare the dose
estimation between Kori Unit 1, the first commercial NPP, and the world's leading
decommissioned NPP. In the future, when the detailed dose for each dismantling
activities is estimated, we expect that this can be used as data to review theextent to which the result value is compared to overseas cases.

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#4 Environmental Radiation Measurement and Assessment#4 Environmental Radiation Measurement and Assessment

Effects of Scenarios and Source Types on Doses for the Reuse
Buildings from the Decommissioning of Nuclear Facilities

Hyung-Woo Seo*, Gang-Woo Ryu
Decommissioning Technology Group/KHNP CRI1
Email: sehywo1727@khnp.co.kr
Keywords: Decommissioning, Building reuse, DCGL
Soil and buildings will be representative of the media remaining after the decommissioning of nuclear facilities is completed. In case of buildings, radioactive
contaminants may be on the surface or may be partially infiltrated and exist in
the depth. For the surface contamination, the effects of either removable or fixedcontamination on receptors shall be assessed. Therefore, this study aims to analyze the impact of surface contamination or volume contamination and building
reuse scenarios when evaluating buildings that will remain after the decommissioning of nuclear facilities.
Building occupancy and renovation scenarios are generally applicable to building
reuse scenarios in the decommissioning of Nuclear Power Plants (NPPs). In thisstudy, the source type was evaluated by setting as the area or volume for the two
scenarios. Since RESRAD-Build code has been widely used as a tool for evaluating building reuse in decommissioning NPPs, we have also applied the scenarios and source type to this code. In addition, in order to use the RESRADBuild code, the input parameter value must be determined. The determination of
the value of the input parameter can be applied differently depending on theapplication scenario, but in this study, the value of the previous study results wasapplied.
When all three radionuclides assumed in this study were under the same conditions, the DCGL of the renovation scenario was rated higher than that of theoccupancy scenario, which means that the dose impact of the renovation was
small. In addition, in the same scenario, the DCGL results from volumetric contamination were higher than DCGL from area sources, which also means that thedose impact of volumetric sources was smaller.
We conducted a preliminary evaluation of comparative parameters using datafrom previous studies in the DCGL calculation for building reuse in the decom-409

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P4 - 32

missioning of NPPs. Evaluation was performed on two scenarios and two source
types. As a result, DCGLs were higher in renovation scenarios than building occupancy and volume than area source. Similarly, it was confirmed that the DCGLof Sr-90 was highly evaluated under the same scenario conditions. In other word,
it was found that the dose effect from radioactive contamination was greater inthe occupancy scenario and greater in area contamination.

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#4 Environmental Radiation Measurement and Assessment#4 Environmental Radiation Measurement and Assessment

Related research on ecological environmental impact assessment
around nuclear fuel reprocessing plant

Gong Wenjing1, Cao Shaofei1,*, Li Jianguo1, Wang xinni1
China Institute for Radiation Protection,Shanxi,Taiyuan ,China1
Email: csf1860@126.com
Keywords: Nuclear fuel reprocessing plant;Emission pathways;Emission
radionuclides;Exposed pathway
Through the investigation of foreign literatures on nuclear fuel reprocessing
plants, this paper summarizes the main emission pathways of radionuclide innuclear fuel reprocessing plants, the main biological exposure pathways in thesurrounding environment of the nuclear fuel reprocessing plants, the monitoringof the main emission radionuclide and the activity concentration in the animals
and plants.Nuclear fuel reprocessing plant emissions of radionuclides has two
kinds of ways the one is a airborne emissions, through the chimney discharged
into the atmosphere, the other is liquid emissions, by a certain distance fromcoast pipe into the sea or rivers.
According to the continuous monitoring data of the effluents from the Tokai
reprocessing plant, the main radionuclides in the airborne effluents are
3H,14C,85Kr,129I, and the main radionuclides in the liquid effluents are
3H,90Sr,137Cs,Pu(α). Due to the presence of backgroundradionuclides in thenatural environment, and the monitoring results are obviously affected by theweather, the data may have short-term fluctuations.
Radionuclides are discharged into the air with airborne effluents and form
aerosols with suspended substances in the air, which are inhaled by animals
through respiration to cause internal exposure.Some of the sediments were
deposited on the soil surface to cause external exposure to the organisms.The
sediments directly fell on the plant leaves and roots, which translocate andtransfer the radionuclides to the primary trophic level.Radionuclides in the air
medium can directly cause external exposure to organisms.
The liquid effluent is discharged into the receiving water in the form of tankdischarge. The short-lived radionuclides will decay out or their activity will
decrease before discharge. However, the aquatic organisms will still be irradiatedby some long-lived radionuclides.It may be absorbed by lower organisms in thewater or concentrated by plants and last enter the food chain of the aquatic
ecosystem.

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P4 - 33

This article mainly summarizes 3H,14C,129I in the terrestrial environment,thethree radionuclides occur pathway, background sources, monitoring methods,
and exposure to biological way after entering the environment were transferred
to plant shift into primary trophic level, according to the formula and the dataobtained from field measurement computing the activity of each radionuclides
concentration in plants and data will be compared with the relevant departmentsof the approved limit.

412

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P5 - 01#5 Radiological Risk Management

Impact of vendor-agnostic deep learning-based CT denoisingtechnique on patient’s radiation dose reduction in a multi-vendor
CT clinical environment

Ahyeong Lee1, Chulkyun Ahn1, Hyeung Woo Nam2, Chang Won Kim3, Kwang
Pyo Kim2, Jong Hyo Kim1,*
a Department of Applied Bioengineering, Graduate School of Convergence
Science and Technology, Seoul National University, Rep. of Korea1, Departmentof Nuclear Engineering, Kyung Hee University, Rep. of Korea2, ClariPi Research,
Seoul, Rep. of Korea3
Email: kimjh@snu.ac.kr
Keywords: deep learning, denoising, Computed tomography, ultra low dose
Introduction
Deep learning technology has been thriving in many fields including medical image diagnosis. Recently, deep learning-based denoising (DLD) techniques for
CT images have been commercialized, and are being deployed to decrease radiation dose to patients and while preserving the image quality. However, commercialized methods provided by the CT scanner manufacturers were only
equipped on their premium CT models, which limits distribution over the world
for the clinical use. On the other hand, software-based vendor-agnostic CT denoising solutions can be distributed without any limitations, allowing lower radiation dose to come to reality in a short period of time over the world. For an
example, ClariCT.AI is a denoising solution which has been proven by many clinical researches to be effective at improving the image quality in low-dose CT
images in a vendor-agnostic way. This study aims to assess the effectivenessof ClariCT.AI on the reduction of patient radiation dose in a multi-vendor CT
clinical environment using Monte Carlo (MC) simulation.
Methods & Materials
This study collected paired CT dataset from a university Hospital, which consistedof routine dose (RD) and ultra-low dose (ULD) CT images of chest and abdomenfrom same patients. In chest dataset, RD set were low dose lung cancer screening examinations (120 kV, 20 mAs), and ULD set were of a tenth of radiation
exposure (120 kV, 2 mAs) to that of RD. The data were acquired from 100 patients(male female ratio, 32:68) using Revolution CT scanner (GE Healthcare). RD CT
images were reconstructed by ASiR-V, which is the vendor’s iterative reconstruction (IR) method, and ULD CT images were reconstructed via filtered back413

Abstracts - Poster

P5 - 01

projection (FBP) and additionally denoised using ClariCT.AI (ClariPi Inc., Seoul,
Korea) which is vendor-agnostic deep learning CT denoising solution. In abdomen dataset, RD set were usual liver contrast enhancement CTs (90 kV, 220 mAs),
and ULD set were of a third of radiation exposure (90 kV, 73 mAs) to that of RD.
The data were acquired from 80 patients (male female ratio, 58:22) using Somatom Force CT scanner (Siemens Healthineer). RD CT images were reconstructedby ADMIRE, which is the vendor’s IR method, and ULD CT images were reconstructed via FBP and denoised using ClariCT.AI. Image quality was evaluated toassess if RD and ULD sets have same level of quality. Expert radiologists evaluated clinical image quality, and furthermore, quantitative image analysis was done.
Three thoracic radiologists evaluated chest structures on RD and ULD CT images
and scored according to 5-point Likert scale on subjective noise, spatial resolution, distortion artifact and overall image quality. As for abdominal CT, 3 abdominal radiologists evaluated liver contour sharpness, liver lesion conspicuity,
pancreas contour sharpness, kidney contour sharpness, image noise and artificial sensation on 5-point Likert scale. For quantitative evaluation, image noise
was measured as the standard deviation of HU on homogeneous ROI, and theimage sharpness was measured using edge rise distance (ERD) acquired bymeasuring 90 % and 10 % of maximum HU at the edge structure. The average
major organ-specific patient radiation dose and total effective dose were calculated using Monte Carlo N-particle (MCNP) code based on CT scan protocol andgeometry information.
Results
In chest CT set, the clinical image quality was 3.02 ±0.19 for RD and 3.89 ±0.19
for ULD exhibiting significant superiority of deep learning-denoised ULD over RD.
Image noise was 47.14 ±18.76 HU for RD and 52.19 ±23.09 HU for ULD showing
comparable noise level, and ERD in pulmonary vessel was 1.46 ±0.17 mm for
RD and 1.30 ±0.17 mm for ULD which meant ULD has superior sharpness compared to RD. In abdomen CT set, the clinical image quality was 3.94 ±0.34 for
RD, and 3.77 ±0.38 for ULD. Image noise was 10.50 ±1.37 HU for RD and 9.94
±1.18 HU for ULD showing comparable level, and ERD in liver vessel was 2.03
±0.08 mm for RD and 1.93 ±0.05 mm for ULD indicating superior sharpness
compared to RD. In the chest CT, effective dose was reduced from 0.9 to 0.1
mSv (88 % reduction) for men, and reduced from 0.8 to 0.1 mSv (87 % reduction)
for women, respectively. In the abdomen CT, effective dose was reduced from
4.5 to 1.5 mSv (67 % reduction) for men, and reduced from 5.0 to 1.6 mSv (68 %reduction) for women, respectively.
Conclusion
Denoised ULD CT images with DLD technique showed equivalent or better image
quality than RD CT images, in a vendor-agnostic way, and reduced patient effective radiation dose by 67 – 90 % compared to RD. The vendor-agnostic DLD
technique has a potential to reduce the patient radiation dose while preserving
CT image quality.
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P5 - 02#5 Radiological Risk Management

Review on Radiation Protection Program during
Decommissioning of NPP

Hye Jin Jung1,*, Hyu Chang Choi, Kwang Ho Jo, Hyung Woo Seo, Jin Won Son
Decommissioning Technology Group, KHNP CRI/Korea1
Email: junghj8113@khnp.co.kr
Keywords: decommissioning radiation protection, internal exposure, plant transition
In contrast to the radiation protection measures in place when the plant is operational, the radiation protection program during decommissioning may be significant and distinctive. The likelihood of airborne radioactivity is often low when theplant is operating, and radiological conditions are typically stable due to system
breaches that are typically short in duration. As components are dismantled andremoved during decommissioning, radiological conditions may vary rapidly andfrequently. In addition, there is a significant increase in the possibility of airborne
radiation due to the removal and segmentation of radioactive systems and components.
The purpose of this study is to review the experiences with Radiation Protection(RP) programs during decommissioning in terms of incidents that have occurred
and the types of controls that are used to avoid incidents, and to find implications
for the decommissioning of Kori-1 Unit and Wolsung-1 Unit in Korea.
As described in the following, decommissioning RP differs from operating RP ina variety of ways. 1) Internal exposure control is being given increased attention
because alpha radionuclides are the main cause of internal exposure concerns,
as all contaminated systems will be breached. 2) Decommissioning RP presents
additional difficulties, including handling a significant amount of highly activated/high dose rate components and components located in high dose areas,
the risk of exposure to hazardous materials, the potential for mixed waste generation, constant plant configuration changes, a significant amount of work hours
under radiological control, and the need for additional RP staff with diverse backgrounds. The Decommissioning RP organization often has a lot more field coverage, pre-job briefing, and planning to do. Examples of increased activities
include the need to develop ALARA plans for particular high-risk tasks, such as
the segmentation and removal of reactor components, and greater engagementof RP throughout all phases of planning for high-risk projects. We also go overthe components of a typical RP program and the kinds of changes that are frequently made to them as a plant transitions from operating to active decommissioning. Additionally, these lessons learned will be applicable to protect workers
even during the decommissioning of Kori-1/Wolsung-1 Unit.
415

Abstracts - Poster

P5 - 03#5 Radiological Risk Management

Current Status Analysis to Optimize Shielding Design Criteria forRadiotherapy Facilities using Linear Accelerators

Sang Hyun Park1,*, Jong Hoon Park1, Beom Suk Ko1, Sung Hoon Ko1
Korea Institute of Nuclear Safety1
Email: shp@kins.re.kr
Keywords: Shielding, Linear accelerators, Optimization, Dose constraint
The general public's dose limit for planned exposure situations is 1 mSv/year.
But it's a nominal value since they may be exposed to multiple radiation sources
and their exposure can not be monitored like radiation workers. Therefore, thedose constraint, set as small fraction of the dose limit has been recommended
to plan radiation protection for the public. It is generally accepted to set the dose
constraint at a range of 10-30% of the annual dose limit (0.1-0.3 mSv/year). In
domestic regulatory framework, the shielding design criteria for nuclear facilities
are based on dose limit, which may require improvement from an optimization
perspective. In this study, for a medical linear accelerator (LINAC) facilities whichis relatively accessible to the public, approaches for optimization were considered in shielding design criteria and practices.
The leakage dose rates outside of shielding barrier in the uncontrolled area wereanalyzed for 230 LINAC facilities currently in operation. For each facility, themaximum value of the shielding assessment results was examined to find out
how many facilities would meet the range of 0.1 to 0.4 mSv/year. The upper limit
was set as one-tenth of the average natural radiation exposure (about 4.0mSv/
year) in Korea.
Only 20% of the facilities satisfied the level of 0.1mSv/year, while 56% met thelevel of 0.4mSv/year. The majority of facilities that exceeded the assessment
value of 0.4mSv/year used 10MV or higher energies or operated at higher workloads than about 10,000 Gy/week. Internationally, in the UK, 0.3mSv/year is applied as the shielding criterion for the uncontrolled area. In the US, where a dose
constraint of 0.25mSv/year is applied, National Council on Radiation Protection
and Measurements (NCRP) recommended a shielding criterion 1.0mSv/year fora LINAC facility in the uncontrolled area. However, due to the inherent conservatism of their recommended shielding assessment methodology, actual exposure
will be at one-fourth of the criterion and would be within the dose constraint.
Based on optimization principle in the domestic regulatory framework, it is possible to apply dose constraints for the shielding design criteria. However, due to416

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P5 - 03the increasing use of workload and the high-energy radiation from LINACs, thiswould be an overly conservative approach asking more resources for facility installation, limiting their operation and potentially conflicting with the philosophyof optimization, ‘as low as reasonably achievable’. An alternative is to maintainthe current criteria based on the dose limit and apply conservative methodologies
in the shielding design and assessment process aiming to achieve a goal that
corresponds to a small fraction of the dose limit.

417

Abstracts - Poster

P5 - 04#5 Radiological Risk Management

Review of the radiation dose of aircraft crew due to cosmic
radiation

Byungmin Lee, Jaeryong YOO*
Korea Institute of Radiological and Medical Sciences, Seoul, Korea1
Email: jerome@kirams.re.kr
Keywords: cosmic radiation, aircraft crew, health effect
Recently, interest in the exposure of aircraft crew due to cosmic radiation has
increased. The ICRP proposes to consider aircraft crew as occupational exposureand recommends continuous management of exposure. Accordingly, countries
such as the United States and France have implemented exposure management
for aircraft crew, and there is a growing need for health effect assessment
regarding cosmic radiation. In Korea, exposure management of aircraft crew iscarried out through the Act on Protective Action Guidelines against Radiation in
the Natural Environment, but no research has been conducted on the healtheffects of aircraft crew. Therefore, in this study, as part of the health effectassessment, the exposure status of domestic and foreign aircraft crew due tocosmic radiation was analyzed. The UNSCEAR presented the radiation dose of
aircraft crew due to cosmic radiation in European countries. The average
effective dose for aircraft crew in seven European countries, including Germany,
the Netherlands, and the United Kingdom, was approximately 1.2 - 2.0 mSv/y.
According to NCRP 160, the average effective dose for aircraft crew in the UnitedStates is about 3.1 mSv/y. In Japan, the average effective dose was found to be
2.2 mSv/y for cabin attendants. In Korea, the NSSC conducts an annual safety
management survey on aircraft crew and provides information such as thenumber of flights, dose status of aircraft crew, and dose evaluation method
through annual reports. The average effective dose for aircraft crew was about
2.1 mSv/y from 2015 to 2019 and 0.9 mSv/y in 2020. The reason for the lower
dose in 2020 is due to a decrease in the number of flights by about 55 - 72%
compared to previous years, as a result of the COVID-19 pandemic. The average
effective dose from 2015 to 2019 was found to be similar to that of othercountries and is less than 5 - 10 mSv/y, which is the reference level of dose foraircraft crew presented by the ICRP. The exposure level of the aircraft crew is at
a low dose. However, in the case of Korea, while health effect assessments fornuclear power plant workers are performed, there is a lack of information on
health effect assessments for aircraft crew. Therefore, to manage the exposure418

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P5 - 04of aircraft crew due to cosmic radiation, it is necessary to conduct research on
reliable dose assessment methods and to analyze dose data. The results of thisstudy can be used as data for the health effect assessment of aircraft crew.
Reference
NCRP. 2019. Ionizing radiation exposure of the population of the United State.
NCRP Report No.160 UNSCEAR. 2010. Sources and effects of ionizing radiation.
UNSCEAR 2008 Report to the general assembly with scientific annexes KINS.
2021. Investigation and analysis of actual state of safety management forradiation in the natural environment, 2020

419

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P5 - 05#5 Radiological Risk Management

A Review on Factors Necessary for Estimating Worker Exposure
Dose for Decommissioning Nuclear Power Plants
Gi-Lim Kim1,*, Hyung-Woo Seo1, Minchul Kim1, Young-il Na1, Jihwan Yu1
KHNP CRI1
Email: todaylover77@khnp.co.kr
Keywords: Worker Exposure, Decommissioning
When decommissioning nuclear power plants, workers are exposed to contaminated buildings, structures, and other materials. Therefore, adjustable factors
could be considered in the work planning phase to align the dose of the worker
below the dose limit and optimize the dose with ALARA. In this study, we would
like to identify what surround factors should be considered to optimize worker
exposure.
First, the radiological status of the workspace is obtained by reflecting the monitoring data or data obtained through radiological characterization. By identifying
the information of the work area, protective measures suitable for the contamination characteristics could be selected. Then, inspect the conditions of theworkspace. When decommissioning, the work area continuously changed according to dismantling work, so it is necessary to monitor how the environment
has changed before the work is implemented. In addition, factors that may affectthe performance of workers’ work, such as scaffolding installation or narrow
space, should be listed.
The next step considers the time and human resource put into the work. The
working time should be appropriately measured by considering the worker’s work
ability or factors that delay the work, while considering the time to perform theactual work. Examples of delaying work include preparing or performing inspection to perform work, and implementing regulatory requirements. The work force
considers both the person who is put into the actual work and the person who
supervises the work. When determining the number of workers, decision makershould calculate as few people as possible to reduce the number of people exposed. However, work that is expected to be exposed to high levels of radiation
should either increase the number or workers or consider a shift system to downlevel the dose of an individual.
Factors such as radiation status, workspace information, time and human resource are quantified and used as data for evaluating exposure doses before
work. It also takes into account the specificity or the work, procedural problems,and other unquantified risks that may affect the radiation work. These systematic
approaches and option identification are described in the work plan at a simple
level and reviewed by the radiation safety management department. In addition,
detailed level could be evaluated by expert groups such as the ALARA committee.
420

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P5 - 06#5 Radiological Risk Management

Comparison of Dose Assessment of Off-link Public According toRadioactive Waste Transportation by Computer Code

Minwoo Kwak1, Geonwoo Son1, Hyeokjae Kim1, Kwangpyo Kim1,*
Department of nuclear engineering, Kyunghee University/Korea1
Email: kpkim@khu.ac.kr
Keywords: RADTRAN, RISKIND, MCNPIn Korea, as a large amount of decommissioning waste is expected to begenerated in the future, the demand for transportation for radioactive waste
disposal will increase. When transporting radioactive waste, it may causeradiation exposure to the off-link public. Therefore, it is necessary to evaluate
the radiation risk prior to transporting radioactive waste. In the case of Korea,
there is no transportation risk assessment code that reflects domestic
characteristics. Therefore, it is necessary to develop the transportation risk
assessment code that reflect domestic characteristics. In this study, prior todevelopment, we evaluated the exposure dose to the off-link public andcompared and analyzed for each overseas computer code. To this end, we
performed (1) assessment code selection, (2) assessment methodology analysis,and (3) exposure situation setting. As assessment codes, RADTRAN, RISKIND
and MCNP computer codes were selected. In the case of RADTRAN, whenevaluating the dose to off-link public, the package is assumed to be a point
source. And to correct for this, the shape factor is applied. RADTRAN basicallydoes not consider the attenuation and build-up of gamma ray for conservativeevaluation. In the case of RISKIND, the dose rate according to the distance isderived based on the statistical dose rate calculated using the Monte Carlo
method for the spent fuel cask. MCNP uses the Monte Carlo method to calculate
results probabilistically by repeating random number generation for various cases.
In this study, we used f5 tally, and nps was set to 2×108. The main exposure
factors in transporting radioactive waste include dose rate, distance, speed, andsize of package. Various exposure situations were set according to the main
exposure factors. In the case of the dose rate, it was set at 0.1 mSv/hr at 1 m.
In additions, we confirmed the exposure dose according to the change indistance, speed, and size of package. As a result of the evaluation, RADTRAN
was generally conservative than RISKIND and MCNP and the results of RISKIND
and MCNP were similar. The conservative reason for RADTRAN is judged to bethat it does not consider the shape factor of RADTRAN and air attenuation and421

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P5 - 06

build-up of gamma rays. And in the case of RADTRAN, the radius of the package
did not affect the exposure dose, because the shape factor of RADTRAN only
affects the length of package if the radius is shorter than the length. All threecomputer codes showed that the exposure dose was inversely proportional to
the speed of vehicle. This is considered to be because the exposure timedecreases in inverse proportion to the speed of vehicle. The results of this study
are expected to be used as basic data when developing radioactive wastetransportation risk assessment codes that reflect domestic characteristics in the
future.
Acknowledgement
This work was supported through the KoFONS using the financial resource
granted by NSSC. (No. 2106041)

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P5 - 07#5 Radiological Risk Management

Assessment of Accident Probability for Overland Transportation of
Spent Nuclear Fuel in Korea

Jooyub Lee1, Jaehyo Kim1, Geehyun Kim1, Kiwon Jang2, ARA GO2, Han-Ki
Jang3,*
Department of Nuclear Engineering, Seoul National University/Korea1,
Department of Radiation Protection and Radioactive Waste Safety,
Korea Institute of Nuclear Safety/Korea2, Department of Future Strategic
Business Bureau, Korea Association for Radiation Application/Korea3
Email: hkjang@ri.or.kr
Keywords: Accident probability, Accident scenario, Sensitivity analysis
In the case of radioactive waste transportation accident conditions, it may cause
damage to the package and, in serious cases, may cause partial or complete
loss of the safety functions of the transportation container, such as shielding andcontainment. To evaluate the radiological impact on human and environment insuch accident condition, the assessment of the impact (risk) is defined as likelihood and magnitude of radiological consequences related to potential transportation accidents and calculated as a multiplication of accident probability andestimated dose.
In order to calculate the accident probability for overland transportation in South
Korea, it is necessary to build a scenario including all accidents that may occur
along the transportation route. We have constructed accident scenarios of trucks
conveying spent fuel and calculated the probability of occurrence for each scenario based on the statistics from the Traffic Accident Analysis System (TAAS) in
South Korea. The frequency of accidents for regular trucks was obtained according to the three different types of roads: highway, general national road, andprovincial road. In addition, occurrence probability of the severity of accidents
was considered according to the damage condition of containing cask to determine the release fraction of certain radioisotope. The frequency of accidents andseverity results according to the overland transportation accident scenarios, including 13 consequence analyses, were derived by applying probabilistic evaluation techniques to provide various information necessary for decision-making.
The sensitivity analysis for each scenario was performed to identify the factorsthat have the greatest influence on the accident probability of potential accidents.
The accident scenarios and the accident probability results developed in thisstudy were compared with the overland accident probability results from United423

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P5 - 07

States Nuclear Regulatory Commission (U.S. NRC). An arbitrary land transportation route was set (Hanbit Nuclear Power Plant – KORAD) for verification, and thevalidity of the accident probability was evaluated.
Acknowledgement
This work was supported by Korea Institute of Nuclear Safety (KINS).

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P5 - 08#5 Radiological Risk Management

Evaluation of Preliminary Risks Derived for the Decommissioningof Nuclear Facilities

Hyung-Woo Seo*, Gi-Lim Kim
Decommissioning Technology Group/KHNP CRI1
Email: sehywo1727@khnp.co.kr
Keywords: Decommissioning, Risk assessment, Decommissioning project
Successful decommissioning of nuclear facilities requires preparation in many
areas such as organization, human resource, cost, and system, but from theperspective of license holder, the well-prepared management of decommissioning project is essential. The License holder in Korea is considered to have enoughexperiences in construction and operation of the facility, and it will need to consider the areas to be reflected in the decommissioning project from the existingprocesses or areas to establish new business flows. In this respect, it will benecessary to perform evaluation of business risks at this stage where the decommissioning project is in the preparation process. Therefore, this study aims
to evaluate and review the preliminary risks derived from the previous study on
the decommissioning project.
As a risk family that can be considered in decommissioning, we can refer to the
IAEA reference. We derived several risks by referring to the current status of domestic decommissioning projects and technical documents related to risks inprevious studies. Therefore, we conducted an evaluation of the derived risks,
which largely fall into 10 of Level 1 categories. In order to evaluate the target risk
group, evaluation items and evaluation criteria must be prepared. In general, thecorresponding risk is evaluated by giving scores with quantitative numbers forboth importance and frequency by referring to the literature.
This study conducted an evaluation on the decommissioning project to be carriedout for the near future in Korea, taking into account the current predictable importance and frequency for preliminary risks. It is expected that there will be risks
to new task that have not yet been systematized, which will require mitigation
measures to be considered in the preparation and progress of dismantling. Since
the results of this study were evaluated for preliminarily set risks, it will be necessary to provide a clear basis for the importance and frequency in future studies.
The results of this study are expected to be used as preliminary evaluation results
for tasks that have never been experienced even though the evaluation was performed with the importance and frequency set for preliminary risks.
425

Abstracts - Poster

P6 - 01#6 Education, Training and Policy in Radiation Safety

Development of Quantitative Assessment System for Evaluation ofTraining Competence in Virtual Reality-based LINAC QA Training

Jongeun Kim1, Wonhyeong Lee2, Yoonsun Chung2,*
Department of Nuclear Engineering, Hanyang University, Seoul, Korea1, Proton
Therapy Center, National Cancer Center, Goyang, Korea1, Department of
Nuclear Engineering, Hanyang University, Seoul, Korea2
Email: ychung@hanyang.ac.kr
Keywords: Quality Assurance, Linear Accelerator, Virtual Reality, Training, Assessment
Purpose:With the advances in diagnostic technology and the increase in cancer incidence
in an aging society, there is a growing demand for cancer treatment and specialized human resources. Medical physicists (MPs), among these specialists,
perform quality assurance (QA) of radiotherapy to ensure the accurate and precise delivery of the prescribed dose to patients. QA is a program that aims todefine the range of acceptable deviations, to detect potential causes of deviation,and develop mechanisms for correction and prevention for radiotherapy. QA requires significant prerequisites, which is why QA training is so important for MPs.
Continued use of facility for patient treatment limits the accessibility and flexibilityof MP's training environments. To complement the constraints of traditional training methods, our team has been developing virtual reality (VR)-based linear accelerator (LINAC) QA training for providing adequate repetitive training. To establish objective evaluation methods for measuring the effectiveness of VR QA
training and confirming the achievement of trainees, it is essential to quantify thetraining data. Therefore, this paper is concerned with the development of a quantitative assessment system (QAS) for evaluation of training competence in VRbased LINAC QA Training.
Methods:
VR simulation environment was constructed using Unity® as physics engine andAutodesk software (Inventor, Fusion 360) as 3D modelling tool. LINAC Daily QA
procedures were selected as VR training contents in consideration of versatility
and frequency of use. Design of training contents was using training scenario
based on consultations with experienced MPs and AAPM task group reports (TG142: Quality assurance of Medical Accelerator, TG TG-198: An Implementation426

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#6 Education, Training and Policy in Radiation Safety

Guide for TG 142 Quality Assurance of Medical Accelerators). The training scenario is divided into three phases: Essential Execution Stage (EES) where the MP
performs the QA protocol in VR; Situational Judgment Stage (SJS) where the MP
practices making informed decisions and taking appropriate actions based on
the AAPM TG-142 recommendation levels; and Scenario Over Stage (SOS) which
evaluates the MP's training performance. QAS is calculated using training dataobtained from training scenario and visualizes competencies of trainee by scoring training data from 0 to 100 in SOS.
Results:
Training of mechanical, safety, and imaging procedure, and their evaluation weremainly focused on the accurate and rapid protocol performances of trainee.
Training data such as each procedure’s Protocol Execution Status (PES), Duration (PED), and Repetition (PER) was used to calculate procedure score (PS).
QAS was designed to reward a high PS with low training penalty introducing stepfunctions and weighted factors. The PS, calculated through QAS using PES and
PED, is stored in PER. The PER with the highest PS among the repeated protocol
executions is displayed as the representative value.
Training of dosimetry procedure and its evaluation was focused on both protocol
performances and the trainee’s ability to make appropriate judgments in specificsituations in a timely manner. PES and PER was only used to evaluate training of
dosimetry procedure. QAS of dosimetry procedure was designed to reward ahigher PS as multiple checks are performed for protocol execution introducing
additional points.
QAS score is designed to reduce the distortion caused by large differences in
PS, and to reward trainees for completing each procedure accurately and efficiently, with higher QAS score given for more successful and timely completionof the entire set of procedures. QAS quantitatively visualizes the competency of
trainees by scoring training data from 0 to 100. The calculated QAS scores are
stored in the QAS database. Through QAS, trainees can assess their own training
achievements and track their progress based on their previous training records.
Additionally, they can compare their own training level to other trainees who havereceived training on the same item by using QAS scores.
Conclusion:
QAS was designed to help trainees assess their competence in VR training for
LINAC Daily QA. Trainees who meet or exceed the training objectives for eachprocedure are awarded higher scores. QAS enables trainees to improve theirscores by minimizing deductions across all procedure training sets. The QAS
database stores and manages the data, which can be used to improve future
training and create effective training plans. Through QAS, trainees can track theirprogress and compare it with others, enhancing their training outcomes.

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Reference
KLEIN, Eric E., et al. Task Group 142 report: Quality assurance of medical accelerators a. Medical physics, 2009, 36.9Part1: 4197-4212. HANLEY, Joseph, etal. AAPM Task Group 198 Report: An implementation guide for TG 142 quality
assurance of medical accelerators. Medical physics, 2021, 48.10: e830-e885.
Lee et al., An approach for establishing digital twin-based VR simulation environment for education and training of radiotherapy professions, 2022, KSMP
Spring Kim et al., Development of training scenario and evaluation standards fordaily QA of LIANC based on digital twin – VR simulation, 2022, KSMP Spring Lee
et al., Constructing scenario-based daily QA contents on digital twin-based VR
simulation for education and training of radiotherapy professions, 2022, KSMP
Fall Kim et al., Development of tutorial scenario for QA training in LINAC VR Simulation, 2022, KSMP Fall Lee et al., Tutorial Development for Digital Twin Simulation in Scenario-based Linear Accelerator Daily QA Training, 2023, KSMP
Spring Kim et al., Quantitative Assessment System of VR QA training competence:
Mechanical and Safety Procedures in LINAC Daily QA, 2023, KSMP Spring428

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P6 - 02#6 Education, Training and Policy in Radiation Safety

Comparison of Radiation Regulatory Systems in Selected
Countries : Focusing on Differences and Adaptations

Euna Lee1, Yoonsun Chung1,*
Department of Nuclear Engineering, Hanyang University, Seoul, Korea1
Email: ychung@hanyang.ac.kr
Keywords: radiation regulation, regulatory body, COVID-19, regulation task division
The application of radiation is beneficial, but it has the potential to be dangerous
if not used properly. Thus, regulations on radiation are essential to ensure its
safe use in our daily lives. These radiation regulatory systems are designed tosuit each country’s environment and are revised as technologies advance or as
the regulatory condition changes such as the pandemic. In this study, we investigated the characteristics of radiation regulatory systems in several abroadcountries, including the USA, Japan, France, Australia, and Canada, and responses to the COVID-19 pandemic, and compared them with Korea.
First of all, one of the differences was the division of duties of the regulatory
bodies responsible for regulating radiation. In Korea, KINS (Korea Institute of
Nuclear Safety) under the NSSC (Nuclear Safety and Security Commission) is incharge of overseeing radiation facilities. In the USA, the agency in charge ofradiation regulation differs depending on the field. For example, NRC (Nuclear
Regulatory Commission), FDA (Food and Drug Administration), Agreement States,and other agencies have specific roles in radiation regulation. In the case ofJapan, under the supervision of the NRA (Nuclear Regulation Authority), four institutions are in charge of various inspections: ‘Registered Inspection Organization’, ‘Registered Periodic Confirmation Organization’, ‘Registered Transport
Method Confirmation Organization’, and ‘Registered Package Confirmation Organization’. Furthermore, in France, ASN (The French Nuclear Safety Authority),
which is similar to NSSC in Korea, divides tasks into the transport and sources
department, ionizing radiation and health department, and regional divisions.
In addition, standards and methods for classifying radiation facilities and radiation sources differ in each country. In Korea, facilities are classified into threelevels according to the risk of the facility, and recently a new regulation has been
introduced to distinguish large radiation generators (RG) such as particle accelerators (more than a few hundred MeV) from small RG. By evaluating radiation
facilities and radiation sources, Australia classified facilities into ‘Hazard andControl levels’ and radiation sources into ‘Groups’ based on a set of criteria.

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Canada categorizes the radiation sources according to their hazard with a ‘Riskinformed oversight regulatory program’. Unlike other countries with complex
standards, Japan simply classifies radiation sources into sealed and unsealed
radiation sources and RG.
Also, each country presented alternative measures in consideration of the
COVID-19 situation in their country. Many countries utilize the Internet or remotesystems to implement regulations. Canada has built a new ‘e-consultation platform’ to assure the convenience of licensees. Australia has also introduced separate digital systems such as PAS (Platforms and Systems) and LIMS (Laboratory
Information Management Systems). However, it seems that Korea tried to implement regulations by supplementing the existing regulatory systems rather thanintroducing new technologies to maintain radiation quality. For example, Korea
continued to enforce regulations such as requiring additional document review
to check the facilities and granting exemption of inspection in some fields.
We reviewed the radiation regulatory systems of some countries and differenceswere analyzed. Each country had its own unique characteristics and methods ofregulating radiation, and it also appeared again in the COVID-19 situation. It isclear that all countries prioritize the safety of utilizing radiation and continuallyadapt their regulations to keep up with changing conditions. Our study emphasizes the importance of continuous adaptation of radiation regulation to changing
conditions, in order to ensure the safe use of radiation in daily life.
Reference
1. Korea Institute of Nuclear Safety (KINS), Foreign Radiation Safety System, BH
Lee. et al, 2019 2. Australian Radiation Protection and Nuclear Safety Agency(ARPANSA), Inspection Manual, 2019.05 3. Korea Institute of Nuclear Safety(KINS), Development of Advanced Radiation Safety Regulation Technology
Based upon Risk-informed and Performance-based Approach, JG Jang. et al,2015 4. Periodic inspection of radioisotopes in each country, KH Lee, 2019 5.
Canadian Nuclear Safety Commission (CNSC), https://www.letstalknuclearsafety.ca/ 6. Australian Radiation Protection and Nuclear Safety Agency(ARPANSA), Annual Report of the Chief Executive Officer of ARPANSA, 2021 7.
Radiation Safety Information System (RASIS), https://rasis.kins.re.kr/

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P6 - 03#6 Education, Training and Policy in Radiation Safety

Analysis of Radiation Monitoring System for Imported Cargo in
Korea, the United States, the United Kingdom, and Japan

Choi Wo Suk1, Kong Tae Young1,*
Department of Chosun University1
Email: tykong@chosun.ac.kr
Keywords: Environmental radiation, Radiation monitoring system, Imported
cargo
The possibility of radioactive contamination in products imported into Korea has
increased since the accident at Fukushima nuclear power plant in Japan. In
addition, recycled scrap metals contaminated with radioactive materials were
imported into Gunsan Port in Korea in 2014. In 2021, a case occurred in which
the radioactivity level of rockfish caught off the coast of Fukushima exceeded
the allowable radioactivity level by 2.7 times. As a result, systematic radiation
monitoring was required for food and daily necessities imported from the nuclear
accident areas. The need for a legal system to support systematic radiation
monitoring was demanded. This study was conducted to review the Korean
radiation monitoring system for imported cargo and compare and analyze theradiation monitoring system for imported cargo in the US, the UK, and Japan asoverseas cases.
Currently, Nuclear Safety and Security Commission in Korea operates a total of
134 radiation monitors at major airports and ports across the country, considering
the number of goods transported and the type of cargo. However, the types ofcargo imported into Korea are diverse and large; therefore, all imported cargo is
not included in the monitoring target of the Nuclear Safety and Security
Commission. Therefore, the radiation monitoring system for Korean imported
cargo is conducted by six government departments, including the Korea Customs
Service, the Ministry of Food and Drug Safety, the Ministry of Environment, theMinistry of Agriculture, Food and Rural Affairs, the Korea Forest Service, and the
Nuclear Safety and Security Commission.
Although the United States, the United Kingdom, and Japan have established andimplemented radiation monitoring systems for imported cargo, the purpose ofmonitoring is different from that of Korea. First, Korea has a radiation monitoring
purpose of protecting the environment and public safety from living radiation, but
the United States has the purpose of performing radiation monitoring on imported
cargo to prevent the threat of terrorism by radioactive materials. Radiation

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P6 - 03monitoring in the UK and Japan is conducted to transport radioactive materials
safely. In addition, in Korea, multiple government departments conduct radiation
monitoring according to legal and monitoring standards; however, in the UnitedStates, the Department of Homeland Security specializes in conducting radiation
monitoring under the Port Safety Act.
This study analyzed the current status of radiation monitoring systems in Korea,
the United States, the United Kingdom, and Japan to compare the purpose andlegal system of radiation monitoring in each country.
Acknowledgement
This research was supported by the Nuclear Safety Research Program throughthe Korea Foundation of Nuclear Safety using the financial resource granted by
the Nuclear Safety and Security Commission of the Republic of Korea (No.
2106024-0323-SB110).

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#6 Education, Training and Policy in Radiation Safety#6 Education, Training and Policy in Radiation Safety

Student Training Program for Radiation Safety in Hanyang
University: HY-Radiedu program and Radiation Safety technology
conference

Euna Lee1, Wonhyeong Lee1, Euidam Kim1, Junik Cho1, Jongeun Kim1,
Chae-eon Kim1, Jiwon Choi1, Yoonsun Chung1,*
Department of Nuclear Engineering, Hanyang University, Seoul/Korea1
Email: ychung@hanyang.ac.kr
Keywords: Student training program, Educational support, Field trip, Overseas
training, Radiation safety techonology conference
Although radiation and its applications have high potential in modern technologies, public interest in the field has been negatively increasing, adversely affecting its prosperity. Recent events, such as the Fukushima disaster and the radon
mattress incident have contributed to this situation. As a result, radiation safety
professionals are necessary, and it is important to nurture nuclear/radiation students to become radiation safety technology professionals.
As part of this, we operated an education and training program for nuclear engineering students, supported by K-CLOUD of KHNP (Korea Hydro & Nuclear
Power). The program, called ‘HY-Radiedu’ (Radiation Education and Training
program in Hanyang University), included educational support, field trips to radiation-related institutions, and overseas training. Additionally, a radiation safety
technology conference will be held this April with six domestic and one international expert presenters.
First, we supported the students with various education programs through our
educational support program. While many radiation-related institutions operate
educational programs, the expenses are often too high for undergraduate andgraduate students to afford. To address this, we provided tuition fees for students
who fulfilled our standards. As of February 2023, we have supported 51 out of
68 applicants. Most of the students (49 out of 68) were interested in code education, such as MCNP, indicating a high demand for code education.
Second, we visited six radiation safety-related institutions with 76 students, including the first heavy-ion cancer therapy center, the first accelerator-based
BNCT facility in Korea, and a nuclear power plant. Students had the opportunity
to listen to detailed explanations about each institution from field experts, which
could help them develop a deeper interest and motivate them to pursue further

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studies or careers in that field. Hereafter, we plan to make up to six more field
trips, including private radiation-related research facilities.
Third, we supported the students in studying abroad for a short term. The specificsite was decided by each student, and we assessed and supported them basedon program validity and their potential. So far, all four students who applied havebeen chosen to be supported by our inner standard. With overseas educationand training support, these students will be able to explore their careers anddevelop their research abilities in international research environments.
Lastly, the radiation safety technology conference is planned to be held at the
Hanyang University Museum conference hall in April 2023. Students will be able
to hear the various research stories and vivid field experiences from radiation
safety technology experts.
Many students have participated in our program for a period of about two years,
and it is clear that they have received meaningful outcomes. With our support,
some students were able to learn new skills, explore their career paths, or taketheir own steps as researchers.
Furthermore, we have positively improved the management system of our program based on the participants’ survey results. As a result, students who haveparticipated in our program once are returning to participate continuously. This
feedback loop is evidence of our program’s positive impact on students.
Acknowledgement
This work was supported by KHNP (Korea Hydro & Nulcear Power) CO., LTD (No.
20-개념-인력-3).

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P7 - 01#7 Radiological Emergency Planning and Preparedness

Review of Establishing EPZ through GIS tools

Bongseok Kim1,*
Nuclear Emergency Preparedness Team, Korea Atomic Energy ResearchInstitue / Korea1
Email: kbs@kaeri.re.kr
Keywords: EPZ, Neclear Emergency, GIS, Protective Action
The establishment of an Emergency Planning Zone (EPZ) for nuclear facilities isa fundamental aspect of emergency planning. An EPZ is typically divided into
two zones: the Precautionary Action Zone (PAZ) and the Urgent Protective Action
Planning Zone (UPZ).
The UPZ is an area where the effects of a nuclear facility accident are closely
monitored and public protective measures are implemented in areas that may beaffected. The PAZ, on the other hand, is an area where preemptive public
protective measures are taken for areas that may be impacted by a nuclear facility
accident. This enables the public to be protected in advance before the effects
of an accident occur.
In Korea, for nuclear power plants, the PAZ is set within a radius of 3 to 5 km,
while the UPZ is set between 20 km and 30 km. For various research reactors
and nuclear facilities, the standard radius is determined based on the size of thepotential risk. For facilities other than nuclear power plants, only the UPZ standard
is presented, with the radius standard ranging from about 0.5 to 5 km.
Establishing an EPZ for nuclear facilities requires consultations between thefacility, local governments, and other stakeholders. Understanding the local
context is crucial in these consultations, and GIS (Geographic Information
System) can serve as a visualization tool to help understand the local context
and produce background data for decision-making.
During consultations for the establishment of an EPZ, it is recommended toconsider various factors such as administrative boundaries, roads, rivers, andpopulation. GIS can be used to visualize this information and provide backgrounddata for decision-making. Through this process, an EPZ can be established, andeffective public protective actions for EPZ residents can be comprehensively
considered.
Overall, establishing an EPZ is an important step in emergency planning for
nuclear facilities, and utilizing GIS to understand the local context can aid in
effective decision-making for public safety.

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P7 - 02#7 Radiological Emergency Planning and Preparedness

Development of GUI Program for Dose Assessment for Skin
Contamination in Radiation Accident

Lee Seung Je1, Kim Han Sung1,*, Yoo Jaeryong1
National Radiation Emergency Medical Center, Korea Institute of Radiologicaland Medical Sciences (KIRAMS), Seoul, Korea 1
Email: hskim87@kirams.re.kr
Keywords: GUI Program, Dose Assessment, Radiation Accident, Monte Carlo
Simulation
In the event of radiation accident, radioactive isotopes may be released into theenvironment, which can lead to contamination of the skin of workers and thegeneral public without dosimeters. When such contamination occurs, dose
assessment is important in terms of emergency response. During the response,
handheld-type contamination monitors are commonly used to measure surface
contamination level. However, determining the personal dose to the human bodymay not be immediately feasible in practice. In the present study, we aredeveloping a user-friendly graphical user interface (GUI) tool that can easily
calculate the absorbed dose and effective dose when the skin is contaminated,
providing an essential resource for the emergency response.
The development of the GUI program can be divided into two main parts:
obtaining the absorbed dose to organs and effective dose for predeterminedcases and developing the GUI interface. To calculate the absorbed dose andeffective dose, the adult mesh-type reference computational phantoms (MRCPs)
and Geant4 Monte Carlo code were employed. The blender software was utilized
to divide the skin of MRCPs. The Qt 6 library and C++ were used in thedevelopment of the GUI program interface.
The input values of our GUI program consist of the degree of surface
contamination measured in Bq cm-2 units, contaminated body parts, types of
nuclides, and estimation period of exposure. Following a method outlined by the
US Environmental Protection Agency (US EPA), the body was divided into nine
regions; the regions were classified as the head, torso, hands, forearms, upperarms, feet, lower legs, and thighs. Taking into account the potential types ofradioactive isotopes that could be released during an accident, four
representative isotopes (i.e., Co-60, I-131, Cs-137, Ir-192) were selected. Inorder to simulate the contaminated areas, a primary particle was uniformly
generated on the skin surface of each divided body region using Geant4 code.

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The relative errors were less than 3% in all simulated results. Since the GUI
program includes all the precalculated results, there is no need for the user towait for dose calculation.
Our GUI program easily and simply converts the degree of surface contamination
into the dose to human body and provides results. We believe that this program
can be an efficient tool for dose assessment when it is challenging to use
equipment such as personal dosimeters.
Acknowledgement
This study was supported by a grant from the Korea Institute of Radiological andMedical Sciences (KIRAMS), funded by the Nuclear Safety and Security
Commission (NSSC), Republic of Korea (No. 50604–2023)

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P7 - 03#7 Radiological Emergency Planning and Preparedness

Derivation of Considerations for Development of Victim Management Procedure in Radiation Emergency

Kihoon KIM*, Jaeryong Yoo
Korea Institute of Radiological & Medical Sciences, Seoul, Korea 1
Email: kh0906@kirams.re.kr
Keywords: Radiation Emergency, Radiological management, Victims
In the case of nuclear accident, people can be injured and exposed by radioactive materials. Victims should receive radiological and medical care, but emergency situation such as nulear accident causes lack of resources, and sufficient
care may not be provided due to lack of resources. In the case of lack of resource,
to distribute resources efficiently, radiological management for radiation victims
is required. However, in Korea, there are few studies on radiological management
for radiation exposed victims in nuclear accident. In this study, considerations todevelop Korean radiological management procedure for exposed victims were
derived. To derive consideration, radiological management procedure for exposed victims of other countries were analyzed. The U.S. HHS(United States Department of Health Human Services) presents radiological and combined injury(medical and radiological) triage criteria in preparation for a nuclear explosion
accident. In these triage criteria, severity of patients was classified based on
resource availability and radiation exposure dose. Resource availability was classified as normal, contingency, fair, and poor condition. IAEA safety report series
no.2 proposed prediction criteria of patient's exposure dose based on the prodromal phase of ARS (Acute Radiation Syndrome) and a patient management
plan. Patients can be classified quickly if the criteria is used. On the other hand,EC(European Commission) presents TIARA (Treatment Initiatives After Radiological Accidents) method to determine appropriate treatment for contaminated victims depends on radiation level. In terms of case of other countries, the following
should be considered to develop Korean radiological management procedure forvictims. Radiological management procedure for victims should consider harmonization between combined injury triage criteria and treatment methods. Management procedure should be made as algorithm in order to be quickly appliedin the emergency situation. Also, management procedure should contain longterm follow-up system due to radiation stochastic effects. In addition, to judge
victim’s symptom quick and clear, it is necessary to prepare a means to determine radiological effect (like clinical symptom classification and lymphocyte
count). It is applicable to develop victim management procedure in radiation
emergency of Korea.
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P7 - 04#7 Radiological Emergency Planning and Preparedness

Development of a Screening Method for Contamination in Urine
Samples during Radiation Emergencies

Seokwon Yoon 1,*, Jaeryong Yoo1
Korea Institute of Radiological and Medical Sciences1
Email: ysw@kirams.re.kr
Keywords: screening, radiation emergency, urianalysis, LSC
In the event of a radiation accident, a bioassay method using excreta samples isavailable for early internal contamination screening. At the site of an accident, ascreening method for total alpha/beta radioactivity in urine samples has beendeveloped and is being used based on previous studies. However, low sensitivity
for beta-emitting radionuclides such as tritium needs to be addressed, and amethod for screening low-level radioactive contamination needs to be developed.
This study focuses on a screening and evaluation method for internal contamination by alpha and beta-emitting radionuclides that can be used in the event ofa radiation accident. A mobile liquid scintillation counter (LSC) was used for a
rapid analysis of the screening procedure. In particular, a screening level basedon a reference dose was evaluated. A screening procedure for tritium contamination, and urine beta analysis was studied. A screening method for gross alphaactivity of urine samples in the laboratory was investigated.
A mobile LSC was used for beta analysis. A screening level based on the energy
range of beta-emitting radionuclides, with Sr-90 as a standard, was evaluated.
Moreover, the screening criteria could be determined by setting a range of thequenching parameter (QP) based on the standard deviation of the measured
value compared to the background value.
The QP range for low-level beta-emitting radionuclides, such as tritium, wasevaluated. A 10-14 working range was obtained depending on the chemical/physical quenching effect of the sample. In addition, a screening level basedon dose was evaluated, and a screening level of 60,000 dpm/mL, corresponding
to 1 mSv for adults within 4 days after intake, was set. For gross alpha screening,the detection limit (DL) was applied. When 20 blank and uncontaminated urine
samples were repeatedly measured, they showed a range of 0.65-0.69 cpm. ThisDL was set as the screening criteria for suspicious contamination and undetectedcases. If the measurement exceeds this screening criteria, visual appraisal of themeasurement spectrum is considered appropriate.

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P7 - 05#7 Radiological Emergency Planning and Preparedness

Radiation exposure dose reduction inradiation emergency medicine

Jaeryong Yoo*, MinSeok Park, Kihoon Kim, Byungmin Lee
Korea Institute of Radiological and Medical Sciences, Seoul, South Korea1
Email: jerome@kirams.re.kr
Keywords: radiation emergency medicine, internal contamination, dose reduction
In the event of a radiological or nuclear emergency, internal/external contamination (or both) and/or exposure by radionuclides can occur. In these circumstances, appropriate radiation emergency medicine is required to mitigate boththe adverse deterministic and probabilistic health effects of victims, including
radiation workers. Medical and/or radiological triage, internal and/or external
dose assessment, decontamination, treatment, long-term follow-up, and psychological support are critical aspects of radiation emergency medicine. These
key elements are necessary for the establishment and implementation of plansfor protective measures and other countermeasures in a radiation emergency. In
response to a radiation emergency, the IAEA recommends the establishment ofan emergency response system and appropriate plans to effectively manage radiation emergencies following response phase. The response phase can be classified into an emergency response, an initial response, and an implementation
phase. Each response phase is related to the radiation emergency medical treatment. The emergency response phase includes triage, emergency medical response, decontamination, and emergency evacuation for external contamination.
The initial response phase includes the conduction of internal/external dose assessment, decontamination of internally contaminated radionuclides, and retrospective dose assessment. In contrast, the implementation phase includes countermeasures such as thyroid gland blocking via stable iodine, evacuation, relocation, and food intake restrictions to prevent or minimize adverse radiation
health effects. In the case of severe internal contamination, an active medical
response is required to reduce the internal dose. The ICRP recommends the useof agents for internal contamination treatment in situations where at least 10
times greater than ALI (annual limit on intake). Representative internal contamination treatment agents include PB (prussian blue) and Ca/Zn-DTPA. The administration of PB can reduce the amount of residual radioactivity in the body by
1/3, and DTPA can increase the excretion rate at which the radioactive material
is released in urine via continuous administration. Such active medical interven-440

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tions in radiation emergency medicine are required to reduce the exposure dose
depending on the situation. The aforementioned measures can reduce the effects
of exposure by 2 to 3 times. However, the protective effect owing to the implementation of protective measures to avoid exposure results in a reduction of the
exposure by approximately 10 to 1000 times. Therefore, to effectively reduce thedose required to protect radiation workers and the general public in a radiation
emergency, a response plan based on appropriate protective measures andemergency medicine must be established.
Acknowledgement
This study was supported by a grant from the Korea Institute of Radiological andMedical Sciences (KIRAMS), funded by the Nuclear Safety and Security Commission (NSSC), Republic of Korea (2104038-0121-CG100)

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P7 - 06#7 Radiological Emergency Planning and Preparedness

A n a l y s is o f M o u n ta in - V a l l e y B re e z e M o rp h o l o g y A f fe c tin g t h e
D iff u s io n a n d D is tr ib u tio n o f P o l l u t a n t s a t a n I n l a n d N u c l e a r F a c il ity

Huang Sha1,*, Zhao Duoxin1, Niu Yanjing1, Zhang Yadong1
China Institute for Radiation Protection， Taiyuan，China1
Email: huangsh_190@163.com
Keywords: Mountain-Valley Breeze; Diffusion; Inland Nuclear Facility; Numerical
Simulation
Topography has a great influence on the dispersion of contaminants in environmental issues. Most of the inland nuclear facility sites in China are located in
mountainous and complex terrain areas, and the complex terrain such as mountains and hills, basins and canyons have a great influence on the dispersion
pattern of pollutants. In case of nuclear leakage or nuclear emergencies, contaminants are affected by the combination of topography, weather systems andlocal valley winds, and their dispersion characteristics exhibit different distribution characteristics. Thus, the study of the characteristics of mountain-valley
breeze at inland nuclear facility and their effects on the dispersion of radioactive
contaminants can provide a technical basis for environmental impact assessment.
The Regional Atmospheric Model System RAMS has a good effect on the simulation of mountain-valley breeze. In this paper, the RAMS model is used to analyze the mountain-valley breeze characteristics of an inland nuclear facility siteby using the ECMWF, a reanalysis data with a resolution of 0.25°, as the initial
meteorological field. A self-developed Lagrangian particle model is also used tosimulate the dispersion trajectory of pollutants in the simulated area. The simulation uses a multi-layer grid nested simulation with a minimum resolution of 20m
in the vertical direction and extends upward to an altitude of 20,000 m. The longwave and short-wave radiation schemes are respectively the Harrington schemeand the Chen scheme.
The simulation results are: for the selected nuclear facility sites, the frequency ofmountain breeze is much higher than that of valley breeze, and the overall frequency of mountain-valley breeze is not high. During the daytime, valley breeze
occur for a short continuous period, usually 1-2 hours, and mostly occurs aroundsunrise, while the frequency of mountain breeze occurrence at night is higher
and the duration is longer, usually more than 4 hours. When the mountain breezeor valley breeze occurs, the wind field in the vertical direction changes significantly.

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It is concluded that: (1) When the system wind is strong, the local wind has lessinfluence on the pollutant diffusion, and the regional wind field shows the characteristics of large wind speed, stable wind direction, consistent wind direction
at high altitude and on the ground, and the pollutant diffuses farther in a shortterm under such wind field conditions, with narrow diffusion range and flattransport trajectory; (2) When the system wind is weak, the mountain-valley
breeze has conditions to occur, and the regional wind field shows a higher windspeed when the mountain breeze or valley breeze occurs, and the wind fieldwithin the boundary layer shows the characteristics of mountain-valley breeze
distribution, and the wind direction at high altitude and on the ground is usually
opposite, under the influence of such wind field, pollutants are easy to gather
near the valley or basin, and not easy to spread to the more distant areas, andhas a greater impact on lower terrain areas.

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P7 - 07#7 Radiological Emergency Planning and Preparedness

Comparison of diffusion simulation results of JRODOS and
NACADOS at a coastal plant site in China

Zhao Duoxin1,*, Huang Sha1, Lv Minghua1, Chen Longquan1
CIRP,Taiyuan,China1
Email: zhaodxxxxx@163.com
Keywords: JRODOS,NACADOS,diffusion models,tracer experiments,comparision
Simulation Comparison of JRODOS and NACADOS diffusion models for field atmospheric tracer experiments at a coastal plant site in China
Duoxin Zhao
(China Institute for Radiation Protection, Taiyuan, China)
The field atmospheric diffusion tracer experiment is a commonly used method
for verifying the effectiveness of migration and diffusion models in the nuclear
accident consequence assessment system. CIRP is implementing a study onverifying the effectiveness of the nuclear accident consequence assessment
system. The study proposes to conduct field atmospheric diffusion tracer experiments in different environmental conditions in China, and use the experimental
data to evaluate the effectiveness of model simulation in the Chinese nuclear
accident consequence assessment system. Currently, the project has conducted
an atmospheric diffusion tracer experiment on a relatively flat coastal plant site.
In order to evaluate the applicability of the three migration and diffusion models
DIPCOT, RIMPUFF, and LASAT in the JRODOS system in China's nuclear power
plant sites, and the effectiveness of the model in the China Nuclear Accident
Consequence Assessment System NACADOS,in this paper, three diffusion models of JRODOS and a three-dimensional particle diffusion model of the NACADOS
were selected to simulate the SF6 tracer experiment conducted at the coastal
plant site in the CIRP effectiveness verification research project.
Same as the currently commonly used effectiveness evaluation methods for atmospheric models, the simulation results were compared with the experimental
results, using various indicators such as FACT,BIAS,MAE,MG,RMSE,VG,FOEX,FB
and other indicators. Finally, the simulation effectiveness of the migration diffusion model in different nuclear accident consequence evaluation systems wasevaluated.For the commonly used fitness indicator FACT, the difference no morethan 10 times between the simulation results and the monitoring values of the
NACADOS is greater than 80%, and the first and second experiments reached
97% and 94% respectively.The difference no more than 5 times is greater than444

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#7 Radiological Emergency Planning and Preparedness

70%, with a maximum of 90%. The overall performance is comparable to that of
the JRODOS model.
For the currently conducted atmospheric diffusion tracing experiments in flat terrain conditions, the performance of the diffusion models in JRODOS and
NACADOS are generally close. After conducting subsequent atmospheric diffusion tracing experiments, the applicability of the model can be further evaluated.

445

Abstracts - Poster

P7 - 08#7 Radiological Emergency Planning and Preparedness

E m e r g e n c y A c t i o n L e v e ls D e v e l o p m e n t S t u d y f o r R e s e a r c h R e a c t o r silje Cho*, Wiho Ha, Chanki Lee, Wontae Hwang, Hyunki Kim
Nuclear Emergency Prepardness Section, Korea Atomic Energy ResearchInstitue, KOREA1
Email: hyilje@kaeri.re.kr
Keywords: Emergency, EAL, Research Reactor
The emergency action level (EAL) has to be established with the hypotheticalaccident scenarios that may occur in the facility, and the background must bepresented and valid. To this end, a review was conducted to develop an emergency action level related to the severe safety function impairment category andradiation level classified by the IAEA.
In this study, in order to develop EAL for research reactor, overseas cases related
to the EAL were reviewed. And in order to develop EAL considering the characteristics of research reactor, review for each accident corresponding to the safetyfunction impairment was performed through the event analysis for accident list
presented in the preliminary safety analysis report and radiological environmental
impact assessment report of research reactor. The derived results for each accident were finally integrated to derive final EAL corresponding to safety function
impairment category. Finally, EALs for 1 type related to loss of cooling capacity,
1 type related to loss of coolant, and 1 type related to fuel damage were derived.
And in order to provide the setting values for develop the EAL of the research
reactor related to high radiation level category, the effluent release rate was evaluated based on the general intervention level for the initial emergency response.
The effective dose equivalent from external exposure due to radioactive plumeand the committed effective dose equivalent due to inhalation in plume wereconsidered. The total effective dose equivalent is the sum of the effective dose
equivalent from external exposure and the committed effective dose equivalentfrom inhalation. The total amount of emission at which the total effective dose
equivalent for 2 days at the site boundary is 10 mSv was evaluated, and setting
values of each nuclides emission amount were calculated as the nuclide ratio of
the maximum hypothetical accident.
This result can be used as a reference when developing the EAL and backgrounds
for research reactors.
Acknowledgement
The study has been conducted as a part of the project of nuclear emergency
preparedness in KAERI.
446

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#7 Radiological Emergency Planning and Preparedness#7 Radiological Emergency Planning and Preparedness

Rapid determination of Pu, Am and Cm in urine sample bytandem quadrupole ICP-MS

Yang Wu1, Maoyi Luo1, Xiongxin Dai1,*, Ni Yuan1, Hongmei Guo1, Jingjing He1
China Institute for Radiation Protection,Taiyuan, China; Shanxi Provincial Key
Laboratory for Translational Nuclear Medicine and Precision Protection,
Taiyuan, China1
Email: daixx@yahoo.com
Keywords: Actinides, urine, tandem quadrupole ICP-MS
In respond to nuclear emergency event, the internal contamination of affected
workers needs to be immediately assessed to guide the follow-up medical
intervention. Rapid bioassay method for measurements of actinides in humanurine samples is the most commonly used for assessment of internal exposure
dose in emergencies.
In this work, a sensitive and rapid simultaneous analysis method of 239Pu, 240Pu,
241Am and 244Cm in small volume (100 mL) urine samples was established byusing ICP-MS with quadrupole in series. The method involves actinide separationfrom a urine matrix by co-precipitation with hydrous titanium oxide (HTiO),
followed by complexation dissolution of precipitation with HNO3+H2O2 andcolumn purification by a TEVA+DGA tandem resin, and finally counting by ICPMS/MS immediately. The overall recoveries were more than 80% for Pu, and morethan 70% for Am/Cm. The interference of urine matrix was suppressed by using
Axial Field Voltage(AFT), and the measurement of Pu, Am and Cm in Q3 Only
mode and MS/MS mode were comparative studied respectively to further reducethe detection limit. The detection limit for the method were estimated to be 0.86
fg for 239Pu, 0.68 fg for 240Pu, 0.74 fg for 241Am and for 0.81 fg 244Cm in 100
mL urine, which meet sensitivity requirement of the emergency monitoring with
100 mSv internal exposure. The method validation results of spiked urine samples
showed good agreement between the measured and the expected values withthe concentrations of 239Pu, 240Pu, 241Am and 244Cm at 1-200 fg. This
demonstrates the feasibility of the developed method for screening type andproportion of contaminated nuclides and assessment of internal exposure dosein emergencies.
The sample pretreatment of this method can be completed within three hours,
and a single analyst can complete the whole analysis of 239Pu, 240Pu, 241Am,
and 244Cm content within 6 hours for a batch of 12 urine samples.

447

Abstracts - Poster

P7 - 10#7 Radiological Emergency Planning and Preparedness

Effects of Neutron, γ-Ray, and Neutron–γ-Ray Mixed Radiations
on Chromosomal Aberrations in Peripheral-Blood Lymphocytes

Zhang Zhongxin1, Bai Jiamei1, Meng Qianqian1, Zhang Ruifeng1, Dang Xuhong1,
Liu Xiaoming1, Yuan Yayi1, Liu Hongyan1
Radiological and Environmental Medicine Department of China Institute for Radiation Protection1
Email:
Keywords: Chromosomal aberrations, neutron–γ-ray mixed radiation, neutron,
relative biological effectiveness, γ-ray
Objective. This study aimed to explore the relative biological effectiveness (RBE)of monoenergetic neutrons and the interactions of mixed radiation by investigating the biological effects of neutron, 137Cs γ-ray, and neutron–γ-ray mixed radiations on chromosomal aberrations in human peripheral-blood lymphocytes.
Methods. Human peripheral-blood lymphocytes were treated with neutrons fromthe Intensive Deuterium and Tritium Neutron Source Science Facility and a
gamma source from the 137Cs γ-ray source of the Chinese Academy of Sciences for follow-up experimental results. The chromosomal aberrations were analyzed in lymphocytes exposed to neutron, 137Cs γ-ray, and mixed radiations,and the RBE values of neutron radiation were calculated. The statistical distribution of chromosomal aberrations was tested using the dispersion index and
Mann–Whitney U test for all doses to conform to a Poisson distribution. The χ2
test was used to compare the yields of dicentric and centromeric rings among
different radiation groups and evaluate the goodness of fit of the dose–response
curve. We used the z test to evaluate the significance of the equation coefficients.
Results. The RBE for 10 dicentrics per 100 cells was 1.31 for neutron radiation.
For 20 dicentrics per 100 cells, the respective value was 2.05. The yields of dicentric and centromeric rings significantly increased in the mixed radiation groupcompared with the 137Cs γ-ray group at the dose of 0.7 Gy.
Conclusion. The considerations of the dose dependence of RBE and the intercellular interactions were indispensable for the precise estimation of the biological effect of neutron treatment on tumors.

448

P7 - 11#7 Radiological Emergency Planning and Preparedness#7 Radiological Emergency Planning and Preparedness

Single-column Chromatography Separation for Simultaneous Determination of Pu, Np, U, Th in Small-volume Urine for Emergency Radioassays

Maoyi LUO*, Xiongxin DAI, Yang WU, Pengyu WANG, Jingjing HE, Yuan NI
China Institute for Radiation Protection, Taiyuan, China1, Shanxi Provincial Key
Laboratory for Translational Nuclear Medicine and Precision Protection, Taiyuan, China2
Email: luo_maoyi@163.com
Keywords: Transuranic nuclides, Analysis method, Urine, Individual monitoringof internal exposure, Emergency radioassays
There is a growing demand for the rapid determination of actinides in biologicalsamples for radiological emergency preparedness. A new method using a single
chromatography separation of TK200 has been developed for the simultaneous
determination of Pu, Np, Th and U in 20 mL of urine by inductively coupled mass
spectrometry (ICP-MS/MS).
In this method, 20 mL of urine samples was acidified to a concentration of 8 M
HNO3. After adding known amounts of radioactive yield tracers (230Th, 242Pu
and 233U) and 0.2 mL of 3 M NaNO2 for valence adjustment, the sample was
loaded onto the TK200 columns for simultaneous radionuclide separation. ThenTh was first striped using 15 mL of 12M HCl + 0.05M NH3OH·HCl, Pu and Np
were second simultaneous eluted using 10 mL of 0.1M HCl + 0.01M HF + 0.05M
NH3OHCl, and U was finally striped using 10 mL of 1.0 M NH4HCO3.
The overall recoveries of Pu, Np, U and Th were higher than 85% for the entire
procedure, and the recovery ratio of Pu/Np were proven steadily with an averagevalue of 1.00 ± 0.05 (n>25). The minimal detectable activities (MDAs) of 20 mLof urine were found to be 0.005 mBq (2.2 fg) for 239Pu, 0.021 mBq (2.5 fg) for
240Pu, 0.00007 mBq (2.7 fg) for 237Np, 0.019 mBq (1.6 ng) for 238U, 0.20 mBq(0.9 pg) for 234U, 0.001 mBq (12.2 pg) for 235U and 0.0019 mBq (0.46 ng) for
232Th by ICP-MS/MS, which was sufficient to meet sensitivity requirements of
the analysis method for emergency monitoring with 100 mSv internal exposure
dose. 20 mL of urine samples spiked with known amounts of 239Pu, 237Np,
238U and 232Th were analyzed to validate the separation procedure, and good
agreement between the expected values and measured results has beenachieved.

449

Abstracts - Poster

P7 - 11

By using vacuum box technology, the total analytical time was minimized andbatch processing allowed analyzing 24 samples in four hours. The ﬂexibility ofthe method also allows for its easy adaptation to various emergency and implement using an automated chromatographic device.
Acknowledgement
This research was financially supported by the National Natural Science Foundation of China (Contract No. 12005197) and Scientific Research Project for
Spent Fuel Reprocessing450

P7 - 12#7 Radiological Emergency Planning and Preparedness#7 Radiological Emergency Planning and Preparedness

Establishment of the Dose-response Calibration Curve for Semiautomatic Analysis of Dicentric Chromosome

Qianqian Meng*, Zhongxin Zhang, Yue Ren, Xiaozhen Li, Yayi Yuan, Jingjie
Wang, Ruifeng Zhang
China Institute for Radiation Protection , Taiyuan, Shanxi, China1
Email: mqqhust@126.com
Keywords: Biological dose estimation, Dose-response curve, Dicentric chromosome, Automatic analysis
The dicentric chromosome analysis(DCA) is an internationally recognized gold
standard for biological dose estimation. The Traditional DCA is time consuming
and laborious making them impractical for mass casualty incidents. Therefore,
automated dicentric analysis becomes the first choice for emergency situations
involving large numbers of people. In this study, we aim to establish a doseresponse relationship curve for semi-automatic analysis of dicentric chromosomes and to study an automated biological dose estimation method.
Peripheral blood lymphocytes of 3 healthy volunteers were irradiated with X-ray,
cultured, harvested, and fixed on slides. The metaphase images of the samples
were collected by the automatic chromosome analysis system, and a dose-response calibration curve was fitted after automatic analysis of dicentric chromosome aberrations and manual verification. Finally, the accuracy of the automatic
dicentric chromosome analysis dose-response relationship curve was verified.
The fitted dose-response calibration curve of semi-automatic analysis of dicentric chromosome was Y =0.0008 (±0.0002) +0.0092(±0.0009) D + 0.0142(±0.0004)
D 2. The error between estimated dose and actual dose in blind sample verification was about 20%, which could be used for biological dose estimation, and theestimation efficiency was higher than manual estimation.
In this study, we successfully established an X-ray-induced semi-automatic
analysis dose-response calibration for dicentric chromosomes, which can reduce manual analysis time several times, and has great application prospects innuclear emergency response to large-scale radiation accidents.

451

Abstracts - Poster

P7 - 13#7 Radiological Emergency Planning and Preparedness

Role of TLR4/NF-κB signaling pathway in hydrogen-rich water
improving acute radiation-induced brain injury in rats

mengya liu1, xiaoman chen2, yong wang2, hui yuan1, huan liu1, jianguo li1, xiujun
qin1,*
Department of Radiology and Environment Medicine, China Institute for Radiation Protection, Taiyuan, Shanxi, China1, School of Forensics, Shanxi Medical
University, Taiyuan, Shanxi, China2
Email: xiujunqin@163.com
Keywords: hydrogen-rich water, ionizing radiation, TLR4/NF-κB signaling pathway, Acute radiation-induced brain injury, cognitive impairment
Purpose: To observe the changes of Toll-like receptor 4（TLR4）/nuclear factorkB（NF-κB）signaling pathway key molecules, NOD-like receptor pyrin domain
3（NLRP3）and cysteinyl aspartate specific proteinase 1（Caspase 1）in theprocess of radiation-induced brain injury induced by ionizing radiation in animal
experiments, and to explore the role of TLR4/NF-κB signaling pathway in the
improvement of acute radiation-induced brain injury induced by hydrogen-richwater.
Methods: Fifteen male Sprague Dawley rats were randomly divided into 3 groups
according to their body weight: control group, irradiated group (IR group) andhydrogen-rich water intervention group (IR+HRW group), with 5 rats in eachgroup. The control group was not irradiated and was given purified water (20ml/kg) by gavage every day. IR group and IR+HRW group received a single 20
Gy whole brain irradiation, 3 days before irradiation, 10 min before irradiation and
30 days after irradiation were given purified water/hydrogen-rich water (20 ml/kg)
by continuous gavage. The general condition of the rats was observed every day,and the body weight were measured on the 7th days after irradiation. On day 7
after irradiation, the animals were dissected. HE staining was used to detect thepathological changes of the hippocampus. The contents of glutathione (GSH),
malondialdehyde (MDA), interleukin-1β (IL-1β) and hydroxyl radical in brain tissue were determined by enzyme-linked immunosorbent assay (ELISA). Quantitative real-time PCR (qRT-PCR) and Western blotting were used to detect mRNAand protein expression levels of TLR4, NF-κB, NLRP3 and Caspase 1 in the hippocampus of rats.
Results: 7 days after irradiation, the rats in the IR group had symptoms such ashead hair removal and salivation, and the symptoms such as head hair removal452

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#7 Radiological Emergency Planning and Preparedness

and salivation of the rats in the IR+HRW group were alleviated compared withthose in the IR group. Compared with IR group, the body weight of animals in
IR+HRW group showed an increasing trend.HE staining showed that comparedwith the IR group, the cells in the IR+HRW group were relatively orderly arranged,with nuclear shrinkage, fragmentation and lysis, and the dark staining of cytoplasm was relatively alleviated. ELISA results showed that GSH activity in IR+HRW
group was increased compared with IR group, and the contents of MDA, hydroxyl
radical and IL-1β were decreased, but the difference was not statistically significant (P>0.05). qRT-PCR results showed that mRNA expression levels of TLR4,
NF-κB, NLRP3 and Caspase 1 in the hippocampus of IR+HRW group were decreased compared with IR group (P0.05).
Conclusion：1. The mRNA and protein expressions of TLR4、NF-κB、NLRP3
and Caspase1 in rats can be caused by ionizing radiation, suggesting that ionizing radiation may cause pyroptosis of brain cells through TLR4/NF-κB signaling
pathway, leading to acute radiation-induced brain injury and cognitive impairment in rats. 2.Hydrogen-rich water can effectively improve acute radiation brain
injury induced by ionizing radiation in rats, and its mechanism may be related toregulating TLR4/NF-κB signaling pathway, inhibiting inflammatory factor, reducing oxidative stress reaction.

453

Abstracts - Poster

P7 - 14#7 Radiological Emergency Planning and Preparedness

Evaluation of the Portable Radiation Portal Monitor’s Usability for
Contamination Screening of Large Groups in a Radiological
Emergency

Young Min Lee1,*, Hyun Ha Lee1, Hyon Sock Chang2, Tae Hyung Lim2
Dept. of Nuclear Emergency Preparedness, Korea Institute of Nuclear Safety,
Korea1, SI Detection, Korea2
Email: k722lym@kins.re.kr
Keywords: PRPM, Usability, Contamination, Screening, Radiological Emergency
Contamination monitoring is an essential activity to respond to a radiologicalemergency. Radiological emergency response organizations operate portable radiation portal monitors (PRPMs) to implement rapid contamination monitoring incase of a nuclear accident. For effective contamination monitoring, it is essential
to understand the characteristics of the PRPMs and use them according to therelated procedure.
The objective of this study is to define the assessment factors based on the
System Assessment and Validation for Emergency Responders (SAVER, U.S. Department of Homeland Security) program and evaluate them in the usability aspect.
The PRPMs’ assessment was conducted by emergency response professionals
who had experience using the PRPMs. Evaluators conducted the assessmentsfor 2 PRPMs, Thermo Scientific’s TPM-903B and Neosis Korea’s NPRM-2VH-L
in all processes to use them from installation and operation to decomposition
with three scenarios. The first is a portal monitor installation scenario, the second
is an equipment familiarization scenario and the last is a contamination screening
scenario. Through these three scenarios, they evaluated the PRPMs with 21 evaluation factors in 4 categories: capability, deployability, maintainability, and usability.
These PRPMs satisfied international requirements such as IEC 62244 and ANSI
N42.35. But there were advantages and disadvantages for each PRPM in theusability aspect. Especially in the case of high background and large group monitoring, it can be some different screening results from the parameters settingsand operation methods of PRPMs.
The results of this study will be applied to understand the PRPM’s characteristics,454

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#7 Radiological Emergency Planning and Preparedness

develop the procedure to use PRPMs in contamination screening of large groups
and improve the usability of PRPMs in a nuclear emergency.
Acknowledgement
This work was supported by the Nuclear Safety Research Program through the
Korea Foundation of Nuclear Safety (KoFons) using the financial resource
granted by the Nuclear Safety and Security Commission (NSSC) of the Republicof Korea (No.2105010).

455

Abstracts - Poster

P8 - 01#8 Radioactive Waste and Current Radiological Issues

Development optimized evaluation technology for work safety
linking location tracking and in situ measurement indecommissioning facilities
Sang Bum Hong1,*
Decommissioning technology development division, Korea Atomic Energy
Research Institute/Korea 1, Decommissioning technology development division,
Korea Atomic Energy Research Institute/Korea 2, Decommissioning technology
development division, Korea Atomic Energy Research Institute/Korea 3,
Decommissioning technology development division, Korea Atomic Energy
Research Institute/Korea 4, Decommissioning technology development division,
Korea Atomic Energy Research Institute/Korea 5, Decommissioning technology
development division, Korea Atomic Energy Research Institute/Korea 6,
Decommissioning technology development division, Korea Atomic Energy
Research Institute/Korea 7, Decommissioning technology development division,
Korea Atomic Energy Research Institute/Korea 8, NEOSIS Korea/Korea 9,
NEOSIS Korea/Korea 10
Email: sbhong@kaeria.re.kr
Keywords: Decommissioning, worker safety, tracking, in situ, measurement
Many countries are moving to immediately dismantle their retired nuclear facilities,
and emerging digital technologies offering significant potential for more effective
project implementation and risk reduction during decommissioning nuclear
facilities. Decommissioning plan of nuclear facilities require the radiologicalcharacterizations and the establishment of a decommissioning process that canensure the safety for worker and environment. By utilizing the rapidly developed
ICT technology, we have developed a technology that can acquire, analyze, anddeliver information from the decommissioning work to ensure the safety duringdecommissioning, optimize the decommissioning process, and actively respondto various decommissioning situations. The established a surveillance systemthat monitors nuclide inventory and radiation dose distribution in decommission
facilities with real time and wireless transmitting technology. To ensure safety
through object detection, tracking, and position estimation of decommissioning
worker, we established a safety evaluation system using stereo cameras and AI(YOLO and EfficientNet) technology. To evaluate performance of the developed
technology, simulated dismantling work to compare the planned and actual doseof dismantling and waste packing scenarios, and evaluated safety through predefined working path with real time tracking of worker location. As a results of456

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#8 Radioactive Waste and Current Radiological Issues

the evaluation, the accuracy of the worker’s location tracking was shown lessthan 1.2m, and worker dose reduced more than 50% in unidentified sources
detection in decommissioning scenarios.
In the future, this study could be useful for improving the worker safety andefficient of dismantling work by linking decommissioning work area with
optimization design in planning and monitoring decommissioning worker.
Acknowledgement
This work was supported by the National Research Foundation of Korea (NRF)
Grant funded by the Ministry of Science and ICT (NRF-2022M2E9A2063396 andRS-2022-00154985)

457

Abstracts - Poster

P8 - 02#8 Radioactive Waste and Current Radiological Issues

Component analysis for solidification after incineration alternative
thermal treatment process of dry active wastes
Ji Hyun YU1,*, Byung Chae LEE, Jong Hoa KIM, Jang Soo SUH
Sae-An Enertech Corporation1
Email: vdnj5308@sae-tec.com
Keywords: dry active waste,volume reduction, thermal treatment, undergrounddisposal, stabilization process
A large amount of dry-activated waste (DAW) is generated during the operationand maintenance of nuclear facilities. And the amount generated of radioactive
waste will increase more and more during decommissioning[1]. Currently, South
Korea has limited space for on-site storage and underground disposal. Therefore,
it is necessary to develop customized technology for safe disposal[3]. The mosteffective way to reduce the volume of DAW is through thermal treatment such as
combustion. However, this method is not currently used due to the generation ofharmful exhaust gases during thermal treatment[2]. Therefore, we conducted thefollowing studies to suppress the generation of harmful exhaust gas and stabilize
underground disposal.
We selected 14 samples identified DAWs produced from actual nuclear power
plants(Kori, Saeul, Wolsong, Hanbit, Hanul) and their respective physicochemical
properties. After the organic element content of each sample was analyzed byElement analysis(EA), the composition and amount of the remaining products
were analyzed by Proximate analysis(PA). At this time, PVC vinyl and shoe covers(vinyl) containing halogen elements were excluded. In addition, Py-GC/MS
analysis was used to analyze the generation of exhaust gas during thermal treatment.
As a result, it was confirmed that most of the samples consisted of organic elements such as C, H, O, etc, and N was detected only in nylon gloves. It was
confirmed that 10 out of 12 DAWs samples were almost nonexistent( ≤ 1.0 %)after thermal treatment. However, the amount of ash remaining in the dust suitand latex gloves was found to be between 13.7 % ~ 13.8 %. As such, it was
confirmed that the volume of DAWs samples was reduced after thermal treatment.
In addition, it was confirmed that dioxin or harmful heavy metals, referred to asenvironmentally harmful substances, were not detected in the exhaust gas analysis.

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On the other hand, for stable underground disposal, it is important issue to stabilize the residual ash after thermal treatments. Therefore, to design the stabilization process, it is necessary to know the amount and chemical properties ofash generated after thermal treatment. For this purpose, ICP-MS analysis was
performed to determine the composition of inorganic compounds in the remaining product after thermal treatment. As a result, it was confirmed that it wascomposed of various metallic elements, and Ca was detected in all samples. Ca
was detected most in dust suits and latex gloves, and it was confirmed that Si
and Al were also included. Other samples also confirmed the presence of inorganic elements such as Ti, Cu, Mg, and Na. Based on these results, we plan toresearch the selection of stabilizers and process design for each sample in the
future.
Acknowledgement
This work was supported by the National Research Foundation of Korea (NRF)
grant funded. It was conducted as part of the Korea Atomic Energy ResearchInstitute's (KAERI) ‘Development and verification of engineering factors for multidisciplinary treatment of decommissioning waste’ collaboration project.
Reference
1. Jin Beak Park, Cheon-Woo Kim, Suk-Hoon Kim, Ju Youl Kim, Determinationof a representative composition of dry active waste from Korean nuclear power
plants, Annals of Nuclear Energy 42 (2012), pp. 89–95 2. H.-C. Yang, Y.-Zun.
Cho, H.-C. Eun, E.-H. Kim, ‘On the Alternative Incineration Technologies for theTreatment of Hazardous Waste’, Korean Chem. Eng. Res., Vol. 45, No. 4, August,
2007, pp. 319-327. 3. Korea Radioactive Waste Management Corporation, 2008.
4. M.-J. Song, C.-W. Kim, Vitrification of Combustible Dry Active Waste Generated from Korean Nuclear Power Plants, Japanese Journal of Health Physics,
Vol.39, p.250, 2004. 5. F. M. M. Suki, A. A. Rashid, Effect of Dispersion Preparation Technique of Calcium Carbonate (CaCO3) Fillers on Mechanical Propertiesof Natural Rubber (NR) Latex Films, AIP Conference Proceedings, Vol.1865,
p.040015, 2007. 6. Kim, C.W., 2011. Consultation Report on Review of a Representative Composition of DAW (in Korean).

459

Abstracts - Poster

P8 - 03#8 Radioactive Waste and Current Radiological Issues

Measurement of Thermal Neutron Fluence Rates of a Proton
Therapy Facility Using a Superconducting Synchrocyclotron- in the Case of IBA ProteusOne at Sapporo Kojinkai Memorial
Hospital, Japan

Hajime NAKAMURA1,*, Hiroshi MATSUMURA1, Go YOSHIDA1, Akihiro TOYODA1,
Eisuke WATANABE1, Kazuyoshi MASUMOTO1, Taichi MIURA1, Hirotaka
NAKAMURA2
Radiation Science Center, High Energy Accelerator Research Organization/
Japan1, Sapporo High Functioning Radiotherapy Center SAFRA, Sapporo
Kojinkai Memorial Hospital/Japan2
Email: hajime.nakamura@kek.jp
Keywords: Activation, Proton Therapy, Thermal neutron
Under contract with the Nuclear Regulation Authority of Japan, we have investigated the activation of various accelerator facilities. These investigations aim todetermine the activation status of the accelerator body, peripheral equipment,
and concrete floor and walls as preparation for the future decommissioning of
the accelerator facilities. A series of investigations were also conducted for proton therapy facilities using a cyclotron (cyclotron type) and synchrotron (synchrotron type) as accelerators. In addition to these accelerator types, we haveinvestigated a proton therapy facility using a synchrocyclotron (synchrocyclotrontype). This presentation will show the measured results of the fluence rate distribution of thermal neutrons inducing activation in the synchrocyclotron-type
proton therapy facility and the estimated activation level for the accelerator room
concrete.
We measured the spatial distribution of thermal neutron fluence rates generated
during treatment operation at the synchrocyclotron-type proton therapy facility
at Sapporo Kojinkai Memorial Hospital (former Hokkaido Ohno Memorial Hospital).
The proton therapy system at Sapporo Kojinkai Memorial Hospital is ProteusOne,
manufactured by IBA, which uses a proton synchrocyclotron with superconducting magnets. The accelerator body is much smaller than the cyclotrons with ordinary electromagnets, such as the proton therapy facilities at the National Cancer Center Hospital East (NCCHE) and Aizawa Hospital. ProteusOne has a single
gantry and a treatment room, making the facility compact. The energy of the460

P8 - 03#8 Radioactive Waste and Current Radiological Issues

accelerated protons is constant, and the proton energy delivered to the patientis adjusted using a degrader. The principle of this energy adjustment is the sameas that of cyclotron-type proton therapy facilities. Although there are two main
methods of proton therapy, the scanning method and the wobbling method, theproton therapy at Sapporo Kojinkai Memorial Hospital adopts the scanning
method.
Thermal neutron fluences were obtained by installing gold activation detectors,
thermoluminescence dosimeters, and thermal neutron track detectors (CR-39)
during the operation and analyzing them after the operation. As previously obtained results, high thermal neutron fluence rates near the degrader were observed in the cyclotron-type proton therapy facilities at the NCCHE and Aizawa
Hospital. On the other hand, at Sapporo Kojinkai Memorial Hospital, the thermal
neutron fluence rates in the room were almost constant. This difference may berelated to the size of the accelerator room at Sapporo Kojinkai Memorial Hospitalis much smaller than those at the NCCHE and Aizawa Hospital. The thermal neutron fluence rates in the gantry and treatment room were extremely low, and theactivation of the room concrete was estimated to be negligibly low compared tothe clearance level. The obtained results are useful for planning the decommissioning of similar facilities.

461

Abstracts - Poster

P8 - 04#8 Radioactive Waste and Current Radiological Issues

Accident Scenario Analysis in Radioactive Waste
Treatment Facility
San Chae*, Won-Hyuk Jang
Korea Atomic Energy Research Institute (KAERI), 34057, Daejeon, Rep. of Korea1
Email: schae@kaeri.re.kr
Keywords: radioactive waste treatment facility, accident, scenario
All nuclear facilities conduct an assessment of the accident to ensure the safetyof the facilities during the licensing acquisition process. The Radioactive Waste
Treatment Facility is designed to ensure safety against all possible accidents.
Nevertheless, the Nuclear Safety Act and related laws assess the impact on workers and the surrounding general public assuming accidents that cannot occur
under normal operating conditions, and ensure that the impact is below thestandards prescribed by the law.
Accident scenario analysis of radioactive waste treatment facilities is performedin the following order; partitioning of facilities, analysis of potential accidents,
derivation of accidents by compartment, analysis of mitigating circumstances,and evaluation of detailed accidents.
Radioactive waste treatment facilities are generally divided into four or five areas;
office area, pre-treatment area, main treatment area, waste storage area, anddrop-off area. Hazards that may occur in these treatment areas include fires andexplosions, and additional waste treatment facilities include vehicles and cranes
used to move radioactive waste. The scenario is derived by considering thesekey potential risks for each area. In addition, the final scenario is derived, including external natural disasters.
The main accident scenarios derived from this accident scenario analysis areevaluated in detail. For example, a vehicle collision with a waste drum in a radioactive waste storage area can cause a fire.
In order to evaluate these accidents radiological, the radiological origin must be
determined, and the following variables must be defined in the source term. The
following equation from DOE Handbook (HDBK) 3010-94, Airborne Release
Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities reflects the calculation for source term:
Q=MAR×DR×ARF×RF×LPF
The factors for each accident are presented in detail within DOE HANDBOOK.
The main contents of each factor are as follows. In the case of a fire, the damage
ratio is about 50%, the scattering rate is 5.0E-04, and it is assumed that theentire amount is inhaled. Using this, if calculated for a radioactive waste treatment facility in the Korea Atomic Energy Research Institute, 8.61E-03 mSv isevaluated at the boundary of the restricted area for 2 hours after the accident.
462

#8 Radioactive Waste and Current Radiological Issues

P8 - 05#8 Radioactive Waste and Current Radiological Issues

Development of Radwaste Management System based on ICT
technology

Jongjin Kim, Tack-Jin Kim, Jin-Woo Lee*KAERI (Korea Atomic Energy Research Institute1
Email: jinwoo@kaeri.re.kr
Keywords: Radioactive waste management, Digital Twin, Augmented reality, Information Management System
The management of the history of radioactive waste is crucial for radioactive
waste management. In addition, there is a demand for the development of adigital system that can display real-time information related to pre-inspection
and disposal suitability evaluation when disposing of radioactive waste. To meetthese requirements and to manage radioactive waste more accurately and transparently, a radioactive waste safety management system based on ICT technology is being developed to meet the needs of radioactive waste managers andregulatory agencies. This paper introduces the technology of using AR technology to check the contents of radioactive waste drums without opening small
waste contents, IoT sensor technology to check the status of drums in radioactive waste storage, and a history management system. Based on the performanceof the prototype DT composed of AR, IoT, and information management systems,
the augmented reality technology allows users to view the shape information andfill ratio of small packaged waste inside the radioactive waste drum, including QR
code management. The basic data of radioactive waste used in augmented reality, sub-packaging, and drums that have been repackaged after processing aremanaged through the information management system. In addition, the IoT sensor signals are transmitted to the digital twin, and the storage information ofradioactive waste stored in the radioactive waste storage, such as temperature,
drum lid opening and closing, and location movement information, are stored
and can be checked in real-time. Through the development of such a system,
we are expecting a safer and more transparent management of radioactive waste.

463

Abstracts - Poster

P8 - 06#8 Radioactive Waste and Current Radiological Issues

A Study on the Packaging Measures for Radioactive Waste of
NPP Decommissioning

Jun Ki Baik1,*, Ji Hwan Yu1, Byong Sop Lim1, Jin Won Son1
Central Research Institute, Korea Hydro & Nuclear Power (KHNP), Daejeon, Korea1
Email: sfbjk95@gmail.com
Keywords: Decommissioning, Radioactive Waste, Packaging, Transportation
When nuclear power plants are dismantled, it is expected that a large amount ofradioactive waste such as metal, concrete, and spent filters will be generated.
The generated wastes are finally packaged through cutting, decontamination,
melting, and measurement, and then delivered to the disposal facility when thesuitability for disposal is confirmed. In order to consider ways to package anddeliver large amounts of radioactive waste generated during dismantling so thatit does not interfere with the dismantling project, this study will present packaging
measures for each type of radioactive waste.
Since domestic nuclear power plants have not been dismantled so far, packaging
measures for each type of radioactive waste have been reviewed through studies
on dismantling overseas nuclear power plants such as the United States andEurope. In the case of overseas nuclear power plants, it was confirmed that radioactive waste was packaged using large square containers made of metal andsoft bags. In addition to the 200L drum used for the packaging of radioactive
waste at domestic operating nuclear power plants, it was confirmed that various
and large-scale containers were used to pack, transport, and dispose of radioactive waste.
In conclusion, a comparative analysis of the packaging of radioactive waste whendismantling overseas nuclear power plants and the development of radioactive
waste packaging containers that can be used for dismantling domestic nuclear
power plants confirmed that various containers are under development in Korea.
It is believed that it will be applicable to the dismantling of Kori Unit 1 andWolsong Unit 1 in the future.

464

P8 - 07

#8 Radioactive Waste and Current Radiological Issues#8 Radioactive Waste and Current Radiological Issues

Study on the Treatment of Radioactive Spent Resin by
Wet Oxidation

Yahui Xi
China Institute for Radiation Protection1
Email:
Keywords: spent resin,wet oxidation,hot test,engineering device
Degrading radioactive spent resin by wet oxidation had advantages of simple
chemical process, mild reaction condition, and little radionuclide overflow. Took
the radioactive spent resin from certain NPP as object, studied the chemical process and device erection about wet oxidation, through which the technologicalparameter of wet oxidation process for different types of radioactive spent resin
was obtained and the engineering demonstration device with capacity of 10kg
per patch was processed. The total running time of the device was 92h for resin
wet oxidation, with results that the average oxidation degradation rate of resin
was 99.770%, showed that the process was good; 99.9% of γ radionuclide wasretained in the oxidation residue, indicating low overflow rate. The volume ofresidual liquid is reduced by 90% after evaporation; The cement solidification forconcentrated residual liquid was carried out, showed that the solidification properties met the requirements of relevant standard, and realized volume reduction
from spent resin to cement solidified body.

465

Abstracts - Poster

P8 - 08#8 Radioactive Waste and Current Radiological Issues

Radioactivation Investigation in SynchrotronType Proton Therapy Facilities

Hiroshi Matsumura1,*, Go Yoshida1, Akihiro Toyoda1, Kazuyoshi Masumoto1,
Hajime Nakamura1, Taichi Miura1, Takeji Sakae2, Naoki Kondo3
Radiation Science Center, KEK/Japan1, Proton Medical Research Center,
University of Tsukuba/Japan2, Medipolis Proton Therapy and Research Center,
Medipolis Medical Research Institute/Japan3
Email: hiroshi.matsumura@kek.jp
Keywords: synchrotron, proton therapy, activation, decommissioning, accelerator
We have conducted activation investigations of various accelerator facilities under contract with the Nuclear Regulation Authority of Japan. This investigation
aims to determine the activation situation of the accelerator itself, peripheral
equipment, and concrete of the floor and walls in preparation for the future decommissioning of accelerator facilities. Decommissioning can be performed
more easily if the non-activated parts are known beforehand.
In the investigation, one of the kinds of accelerator facilities was a proton therapy
facility. There are two types of proton therapy facilities: cyclotron type using a
cyclotron as a proton accelerator and synchrotron type using a synchrotron. The
significant difference between the two types is that the cyclotron type has a constant acceleration energy and the energy of the protons irradiated to the patientis adjusted by a degrader. In contrast, the synchrotron type can adjust the acceleration energy itself. The effect of these differences on activation is interesting.
We have already reported the investigated results of the activation characteristics
in cyclotron-type proton therapy facilities (National Cancer Center Hospital East
and Aizawa Hospital), which are currently in operation. Neutrons were generated
by beam loss around the degrader and strongly induced activation in the cyclotron-type proton therapy facilities. The activation level was found to be close to
the clearance level in the concrete floor around the degrader.
In this study, we investigated the activation characteristics in synchrotron-type
proton therapy facilities at Proton Medical Research Center at the University ofTsukuba (Univ. of Tsukuba) and Medipolis Proton Therapy and Research Center(Medipolis). The synchrotron at the Univ. of Tsukuba has two treatment rooms
with a maximum proton acceleration energy of 250 MeV. The synchrotron at the
Medipolis has three treatment rooms with a maximum proton acceleration energyof 235 MeV.

466

P8 - 08

#8 Radioactive Waste and Current Radiological Issues

Activation surveys of the beamlines were performed using NaI scintillation survey
meters to measure dose rates and LaBr3 detectors to measure gamma-ray spectra. The activation of the concrete floor was estimated from the measured thermal
neutron fluence by placing gold activation detectors and thermoluminescence
detectors on the floor during operation. The floor was also measured with Ge andCeBr3 detectors directly on the floor immediately after the operation, and activation of the floor was estimated from the Na-24 activity. The measurement results found that the synchrotron-type proton therapy facilities are unlikely to be
activated due to low beam loss in the accelerator room. It was concluded thatthe activation of the floor concrete was negligible compared to the clearance
level and that it can be concluded that the concrete will not be activated at thetime of decommissioning.

467

Abstracts - Poster

P8 - 09#8 Radioactive Waste and Current Radiological Issues

Analysis of Low-Level Radioactive Concrete Waste using Gamma
Spectroscopy applying with Efficiency Transfer

Kyunghun Jung*, Minyoung Jung, Woojin Jo, Jung Bo Yoo
Korea Atomic Energy Research Institute1
Email: kyunghunjung@kaeri.re.kr
Keywords: Low-Level Radioactive Concrete Waste, Gamma Spectroscopy, Efficiency Transfer
Concrete is a major building material for nuclear facilities, and some of it is contaminated or radiated during various activities such as operation and maintenance of the facility. Therefore, it is the most common form of radioactive waste
along with metal waste when nuclear facilities are dismantled. Of the total waste
generated from the Connecticut Yankee and Maine Yankee nuclear power plants
in the US, concrete waste accounted for 83.5% of the total in Connecticut Yankee
and 52% in the case of Maine Yankee. It is known that most of the concrete
generated from the dismantling of nuclear facilities is generated from biologicalshields and is also subject to clearance. However, radioactive concentrations of
gamma nuclides such as Co-58, Co-60, Cs-137, and Ce-144 must be analyzed
in order to dispose of low- and intermediate-level radioactive concrete waste.
Gamma spectroscopy using HPGe is generally used to obtain the radioactive
concentration of gamma nuclides. However, summing of signals occurs when
two or more gamma rays are emitted from the same source due to the nature of
the detector in gamma spectroscopy. In particular, the summing of gamma raysfrom nuclides such as Co-60 or Y-88 that decay sequentially is called true coincidence summing (TCS). Since the full energy peak efficiency to be originallydetected decreases when TCS occurs, it must be corrected in order to accurately
evaluate radioactivity of sample. Therefore, several methodologies and computercodes have been developed for TCS correction.
The radioactive waste radiochemical analysis section of KAERI conducts nuclide
analysis for final disposal of low- and intermediate-level concrete waste. Sincea large number of samples must be analyzed in the facility, the analysis method
must satisfy accuracy and speed at the same time. In this study, we report theresults of applying efficiency transfer, which seems to satisfy these requirements
among the TCS correction methodologies, to gamma spectroscopy.

468

P8 - 10#8 Radioactive Waste and Current Radiological Issues#8 Radioactive Waste and Current Radiological Issues

Preliminary study of the effect of dominant Co-60 in gamma
spectrum of radioactive waste

Sooyeon Lim1,*, Kahee Jeong1, Ji-eun Park1, Wook Sohn1
Central Research Institute of KHNP, Korea1
Email: Sooyeonlim@khnp.co.kr
Keywords: radioacitve waste, nuclide inventory, MDA
In accordance with the regulations of the Nuclear Safety and Security Commission (NSSC) for the disposal of low- and intermediate-low-level radioactive
waste, the activity concentration of essential nuclides in radioactive waste must
be determined. Additionally, the ability to assess 1% of the concentration limit of
low-level radioactive waste is necessary to determine the nuclide inventory inradioactive waste generated by domestic nuclear power plants. Therefore, it is
crucial to obtain a proper level of minimum detectable activity (MDA) when analyzing nuclides.
Unlike other samples such as environmental or food samples, the concentrationof Co-60 in radioactive waste is relatively higher than that of other nuclides.
When high-energy gamma rays (1173 keV and 1332 keV) emitted by Co-60 increase the Compton continuum generated by the Compton scattering effect on
the detector, it causes an increase in MDA in the relatively low-energy region (To analyze the MDA change of Cs-137 as the radioactive concentration of Co60 is differentiated, a simulated sample containing both Co-60 and Cs-137 wasused. The concentration of Co-60 was much higher than that of Cs-137 so thatthe difference could be maximized. Then, the concentration of Co-60 was reduced using a chemical separation method, and gamma spectroscopy was performed using a High Purity Germanium detector (HPGe, Canberra). The reductionof Co-60 and the MDA change of Cs-137 were assessed simultaneously as aresult of gamma spectrometry.
An extraction chromatography method using an anion exchange resin was conducted as a chemical separation method to separate cobalt. Gamma spectroscopy analysis was conducted using HPGe for the remaining elution solution after
cobalt was removed during the extraction chromatography rinsing process. The
eluted solution was put into a 20 ml vial and measured for 10,000 seconds. The
detector was first measured for the same amount of time using a certified reference material(CRM) that contains mixed gamma emitting sources with the same
geometry, and then calibrated for channel-energy and channel-efficiency.

469

Abstracts - Poster

P8 - 11#8 Radioactive Waste and Current Radiological Issues

Dose Assessment for Regulatory Clearance of Organic Liquid
Radioactive Waste at KAERIKi Baek SHIN1, Hyun Myoung JUNG1, Dong Su KIM1, Tack Jin KIM1, Gyu Seong CHO2,*
Korea Atomic Energy Research Institute1, Korea Advanced Institute of Scienceand Technology2
Email: gscho@kaist.ac.kr
Keywords: Dose Assessment, Regulatory Clearance, Organic Liquid Radioactive
Waste, GASPAR code, RESRAD code
The Korea Atomic Energy Research Institute (KAERI) has generated various typesof organic liquid radioactive waste through its nuclear facilities. The majority ofthis waste exhibits low levels of radioactivity. Due to the lack of an establishedtreatment method for organic liquid radioactive waste, it is temporarily stored at
the Radioactive Waste Treatment Facility (RWTF) at KAERI. This is in contrast to
inorganic liquid radioactive waste, which can be treated using recognized techniques. Currently, regulatory clearance is the only option for waste that meets theAtomic Energy Act of Korea. In this study, the dose assessment method andprocess for regulatory clearance of organic liquid radioactive waste stored in
RWTF at KAERI were described.
As a method of regulatory clearance for organic liquid radioactive waste, there is
the option of incineration followed by landfilling, and dose assessments must be
performed sequentially. In the case of incineration, the exposure dose must be
calculated taking into account the impact of radioactive nuclides emitted during
the combustion of radioactive waste. In the case of landfilling, the radiation exposure of residents on top of radioactively contaminated soil must be evaluated.
To evaluate this, the GASPAR and RESRAD-onsite codes were used. For eachcode, dose assessment is carried out based on scenarios that use parametervalues set by the user, such as radionuclides, concentrations, site specifics, andlocation. The codes have default values but users can modify them. The codes
estimate doses resulting from inhalation, ingestion, and external exposure.
Dose assessment by the codes was performed on organic liquid radioactive
waste with a very low possibility of radioactive contamination stored in RWTF at
KAERI. As a result, it was confirmed that some wastes met the regulatory clearance standards. According to the Act, radioactive waste can be cleared by incineration, landfilling and recycling if it meets the criteria of less than 10 uSv/yfor individual dose and 1 person ∙ Sv/y for collective dose.
Based on this, the regulatory clearance is expected to be possible, and if permission from the regulatory body will be obtained, it will be the first regulatory
clearance of organic liquid radioactive waste at KAERI.
470

P8 - 12#8 Radioactive Waste and Current Radiological Issues#8 Radioactive Waste and Current Radiological Issues

Degumming test of radioactive filter element

Yongtai Yuan, Wei Xu*, Bowen Zheng, Peng Wang, Yu Zhang, Zhijie Bai,
Jiashen Ruan
China Institute for Radiation Protection，China1
Email: 13700541402@163.com
Keywords: radioactive filter element,Electromagnetic induction heating,Degumming
With the increase of the number of nuclear power plants in China, the amount ofradioactive waste generated by nuclear power plants is also increasing. Especially, there are a large number of nuclear grade high efficiency filters (HEPA).
Through investigation and investigation, it is found that there are 346 high efficiency filters in Ningde 1 and 2 units alone, which need to be replaced every year
according to the operation conditions. It will produce up to 37.61m3 of HEPA
radioactive waste per year, occupying storage resources and bringing radiation
risks. In order to realize clean decontamination and resource recovery, it is necessary to disassemble the filter element and filter frame, and the filter element
can be compressed or burned for capacity reduction treatment. After scraping
and decontamination, the metal resources can be recycled. However, the sealantat the connection between the filter element and the frame is difficult to deal withand is also dangerous. Due to the tearing and scattering of the filter element in
the operating environment, fine glass fibers will float in the environment, seriously
polluting the working environment and causing radiation hazards. This paper
mainly studies the colloid breaking temperature between the filter element andthe frame, in order to reduce the workload of the staff, reduce the radiation hazard of the staff.
The fragments with colloid and thick colloid were selected in the debris bucket
for testing, and the fragments were heated by high-frequency induction heater.
Four groups of samples were taken for testing, and the separation of colloid andframe, temperature range, and smoke generation conditions were determined.
Below 160℃, such as heating to 140℃, 130℃ and other temperature conditions,a small part of the colloid embrittleation, crisp, compared with the unheated fragments although there is an effect, but the effect is not as good as 160℃ heating
effect;About 160℃, after heating, a small part of the colloid still has a certain
elasticity, into strips or small sheets, most of the embrittling, crisp, easy toshovel;Between 160℃ and 170℃, control the heating time, can do not smoke

471

Abstracts - Poster

P8 - 12and slight smoke, and peeling effect is good, the glue on the bonding surface isbasically crisp and brittle, easy to shovel;170℃ and above, basically smoke, thelonger the time, the more serious the smoke, peeling effect is better.
High frequency induction heating filter element is found to be of great help toglue removal, most of the colloid after heating, crisp, brittle, and due to the colloid and frame physical property difference is large, once heated, colloid andframe separation becomes easy. Therefore, it is feasible to use high frequency
induction heating method as a means of glue removal, which can reduce theworkload of the staff and reduce the radiation hazard.

472

P8 - 13#8 Radioactive Waste and Current Radiological Issues#8 Radioactive Waste and Current Radiological Issues

Recommendations for the Supervision of Mobile Radioactive
Waste Incinerator
Su Ziqiang, Dong Yuyang, Chen jiachen, Lian Bing
China Institute for Radiation Protection1
Email:
Keywords: Radioactive Waste Incinerator；Mobile；Supervision
Background: With the development of nuclear energy application and nuclear
technology utilization, more and more radioactive combustible waste is produced
and scattered all over the country. Stationary incineration devices have high timeand expense costs and also involve the transportation of radioactive materials
when disposing of radioactive wastes from other facilities. Mobile radioactive
waste incinerators will solve this problem, but they pose regulatory challenges
due to their mobility.
Purpose: Ministry of Ecology and Environment of China had incorporated incineration facilities into the management of emission permits and issued a series ofstandard guidelines to supervise them. Except for the field of the marine incinerator, no regulations and standards for mobile incinerators have been issued inother fields in China. The pollutant emission control of Marine incinerators is
stipulated in GB/T 10836-2008, and the flue gas emission measurement is regulated in GB/T30505-2014. The CO in the smoke emission is monitored, and theremaining parameters are controlled through the performance parameters of theincinerator.
Conclusions: 1. Formulate corresponding pollutant emission control standards
with reference to Non-radioactive field and standardize their incineration activities;
2. For incineration activities before the establishment of standards, if the exhaustgas system of the mobile incineration device can be connected to the waste gas
treatment system of the waste generating facility, a summary of emissions should
be made after the incineration activities, and reported to the regulatory authorities;
3. If the exhaust gas system cannot be connected to the waste gas treatment
system of the waste generating facility, When the amount of radioactive combustible waste or the total activity is small, the emission of pollutants can becontrolled through the performance parameter control of incineration device,
which has certain reference value. When the amount of combustible waste or the
total activity is large, the method of performance parameter control will lose its
reference value. It should be carefully considered and the activity should be approved by the regulatory authority.
473

Abstracts - Poster

P8 - 14#8 Radioactive Waste and Current Radiological Issues

In-situ evaluation of neutron flux using activated reaction of55Mn(n,γ)56Mn, 23Na(n,γ)24Na on the component of a tandem
accelerator

Go YOSHIDA1,*, Hiroshi MATSUMURA1, Hajime NAKAMURA1, Taichi MIURA1,
Akihiro TOYODA1, Kazuyoshi MASUMOTO1, Takayuki NAKABAYASHI2, Makoto
MATSUDA3High Energy Accelerator Research Organization (KEK), Tsukuba-Ibaraki, Japan1, Japan Radioisotope Association, Tokyo, Japan2, Japan Atomic Energy
Agency, Tokai-Ibaraki, Japan3
Email: yoshigo@post.kek.jp
Keywords: Activation, Neutron flux, Tandem accelerator
Accelerator activation become apparent as a radioactive waste problem when thefacility is decommissioned. Mainly neutrons generated by beam-loss during accelerator operation triggers activation. Evaluation where and how much neutrons
are generated will greatly contribute to shielding design and waste reduction.
Dosimeters such as solid-track detector (CR-39) and thermos-luminescence
dosimeter (TLDs), and/or activation detectors such as gold and indium have often been used to measure neutrons. Although, they must be installed before operation, and the number of data cannot be increased after operation. Increasing
the number of installation points increases the cost. Therefore, we have developed a method to estimate the neutrons generated during operation by measuring short-lived radionuclides produced in the accelerator components throughneutron capture reactions as activation detectors.
The 20MV tandem accelerator of Japan Atomic Energy Agency (JAEA), the largest
in Japan, was selected for the target facility. The facility consists of seven floors
above ground and two floors below ground. Pressure vessel contains the tandem
accelerator is installed from second to six floors vertically. Maximum 36 MeV
proton beam is available in this facility. We installed TLDs and gold-foils at theentire beamline including the pressure vessel and concrete floors and walls where
beam loss was expected. After the accelerator operation, installed dosimeters
were withdrawn and deduced neutron flux during the accelerator operation. In
parallel, γ-ray spectrometry with a LaBr3-detector and a Ge-detector was performed at the same as the dosimeters within 5 minutes per points. Neutron flux
during operation was estimated from the activity of 56Mn and 24Na produced in474

P8 - 14#8 Radioactive Waste and Current Radiological Issues

the pressure-vessel or concrete building. Thermal neutron flux around the pressure-vessel estimated from 56Mn activity was 1 × 10^2 cm^−2·s^−1 to 8 ×
10^2 cm^−2·s^−1, and they were consistent with the result of TLDs and gold
foils. The value less than 1 × 10^2 cm^−2·s^−1 could not be determined due toinsufficient statistics. Though the highest value was order of 10^4 cm^−2·s^−1,
we concluded that there is no risk of activation in this facility considering theaccumulation of 60Co and 152Eu activity.

475

Abstracts - Poster

P8 - 15#8 Radioactive Waste and Current Radiological Issues

Derivation the activity of 129I limit for near surface disposal

Yangjun Zhao1,*, Jie Yang1
China Institution for Protection,Shanxi Taiyuan, 0300061
Email: zyj218@163.com
Keywords: Disposal site, Activity limit, Waste disposal, 129I
Radioactive waste is an inevitable companion in the process of nuclear energy
utilization. With the rapid development of nuclear energy utilization, a large
amount of low-level radioactive solid waste will be produced, actively and prudently dispose of these low-level radioactive solid wastes is a powerful guarantee
to ensure nuclear and radiation safety. Currently, the problem of lagging behindin the disposal of radioactive waste is becoming more and more serious. The
total amount of radioactive waste temporarily stored in the nuclear power plant
waste has approached or exceeded its original design capacity, and even exceeded the requirement for temporary storage of no more than 5 years. The construction and scale of disposal sites are far behind the development of nuclear
power. Therefore, the disposal of low-level radioactive solid waste has become
one of the main dilemmas faced in the operation of nuclear power plants. In
November 2021, the Central Committee of the Communist Party of China and theState Council pointed out in the opinions on in-depth pollution prevention andcontrol, it is necessary to actively and steadily promote the disposal of radioactive and companion radioactive waste.
The 129I nuclide has a very long half-life (T1/2=1.57×107a), which is not easy
to be adsorbed by the lithosphere but is easy to enter the biosphere throughgroundwater and then enter the food chain. The public is exposed to the radiation
by ingesting food contaminated with this nuclide. Current research adopted theactivity limit derivation method recommended by the International Atomic Energy
Agency (IAEA) to study and deduce the activity limit of 129I in the received anddisposed waste at Longhe Disposal Site, which could provide a basis for theenvironmental impact of receiving 129I-containing waste in the Longhe Disposal
Site.
According to the environment conditions and hydrogeological data surrounding
the Longhe near-surface disposal site for the low-level radioactive solid waste,
by developing and determining the normal evolution, drilling scene and postdrilling scene of the disposal site, the radiation dose to the public generated fromeach scene was calculated by using the latest method recommended by the IAEA,
which could further derive the upper limit value of 129I activity in the wastes
received under each scene of the disposal site when met the requirement of does
criteria. Finally, the smallest upper limit value was selected as the limit of 129I
activity in the waste received by the Longhe near-surface disposal site.
476

P8 - 16

#8 Radioactive Waste and Current Radiological Issues#8 Radioactive Waste and Current Radiological Issues

Comparison of Cost about Equipment for Spent Nuclear FuelManagement by Dry Storage Method in Korea

Hyeok Jae Kim1, Shin Dong Lee1, Min Woo Kwak1, Kwang Pyo Kim1,*
Department of Nuclear Engineering, Kyung Hee University, Korea1
Email: kpkim@khu.ac.kr
Keywords: Equipment for SNF Management, Cost, Dry Storage
According to the spent nuclear fuel (SNF) management policy roadmap presented in Korea, on-site temporary storage facility will be operated until the construction and operation of interim storage facility, and each facility will select adry storage method. Detailed operation method of each facility will be selected
in reflecting of engineering safety and communication with residents in the future[1]. Therefore, various dry storage scenarios can be assumed in the future, andconsideration about cost is required when selecting the optimal scenario. In thisstudy, as a part of previous research for selecting optimal scenario, we compared
cost about equipment for SNF management by hypothetical dry storage scenario.
To compare the equipment cost by scenario based on dry storage method, weinvestigated foreign and domestic SNF life cycle cost estimation cases of AREVA,
Energy Solution, etc. Based on the analyses, we selected equipment for dry storage of SNF to be evaluated, and established a cost classification system bydividing them into hardware cost, loading/unloading cost, and transportation cost.
With this cost classification system, we established equipment cost estimation
system for each phase of the SNF management. Finally, we estimated the costof equipment for hypothetical SNF dry storage management scenarios, and compare results by scenarios [2]. Four hypothetical scenarios were assumed, reflecting the policies about energy and radioactive waste in Korea. It was assumed
that on-site temporary storage facility will be operated based on metal cask,
according to the recent dry storage facility construction plan for Kori nuclear
power plant (NPP). As a result of comparison of equipment cost, it was foundthat among the three scenarios, the lowest equipment cost was incurred in the
scenario where canister type metal cask is utilized in both temporary and intermediate storage phases. In the other two scenarios, equipment cost was 17%
higher than that scenario on average. In all scenarios, hardware cost was dominant in total cost of equipment for SNF management. Hardware cost differedsignificantly by scenario, whereas the loading/unloading cost and transportation
cost were estimated identically in all scenario. The results of this study can beused as basic data when selecting an optimal SNF management scenario.
477

Abstracts - Poster

P8 - 16

Acknowledgement
This work was supported by the Korea Institute of Energy Technology Evaluation
and Planning (KETEP) and the Ministry of Trade, Industry & Energy (MOTIE) of
the Republic of Korea (No. 2021171020001B).
Reference[1] MOTIE, The 2nd Basic Plan for High-Level Radioactive Waste Management,
2021. [2] KORAD, The Implementation Plan for Radioactive Waste Management,
2022.

478

#8 Radioactive Waste and Current Radiological Issues

P8 - 17#8 Radioactive Waste and Current Radiological Issues

Analysis of Recyclability of Concrete Wastes Generated from Nuclear Facilities during Decommissioning

Ji Ung Kim1, Tae Young Kong2,*
Department of Nuclear Engineering, Chosun University, Gwangju, Korea1, Department of Nuclear Engineering, Chosun University, Gwangju, Korea2
Email: tykong@chosun.ac.kr
Keywords: concrete waste, recycling, circulating coarse aggregate, circulating
fine aggregate after
Nuclear facility buildings are made of reinforced concrete, which generates alarge amount of concrete waste when nuclear facilities are dismantled. In thecase of concrete waste, the disposal cost continues to increase. Thus, a technology for recycling a large amount of
concrete waste is emerging as an important technology. Recycling concrete
waste is a good direction not only from an economic point of view but also froman environmental point of view. Therefore, appropriate recycling technology is
indispensable to reduce the amount of concrete waste generated in large quantities when dismantling nuclear facilities in the future.
This study investigated ways to recycle concrete waste. It was found that mostradionuclides were deposited on the concrete surface. In addition, 60Co, 63Ni,
and 137Cs accounted for 99% of the nuclides deposited within 1 mm. This indicated that the uncontaminated parts could be recycled after separating the contaminated concrete parts. The contaminated concrete parts could also be reproduced as circulating coarse aggregate and circulating fine aggregate after crushing, grinding, and decontamination. The recycled coarse aggregate and the recycled fine aggregate could be used as concrete and cement mortar, respectively.
These data are expected to help recycle concrete waste from nuclear facilities
during decommissioning in the future.
Acknowledgement
This research was supported by the Nuclear Safety Research Program throughthe Korea Foundation of Nuclear Safety using the financial resource granted by
the Nuclear Safety and Security Commission of the Republic of Korea (No.
2106024-0323-SB110).

479

Abstracts - Poster

P8 - 18#8 Radioactive Waste and Current Radiological Issues

Radiochemical analysis for difficult-to-measure radionuclides inradioactive samples

Myung Ho Lee1
Young Jae Maeng, KRIST, South Korea1, Mi Joung Lim, KRIST, South Korea2,
Choon Sung Yoo, KRIST, South Korea3, Byoung Chul Kim, KRIST, South Korea4
Email:
Keywords: Sequential Radiochemical Separation, Radioactive Samples, Pu, Am,
Tc, Sr, Nb, Fe and Ni Isotopes
This study presents rapid and quantitative sequential radiochemical separation
method for Pu, Am, Tc, Sr, Nb, Fe and Ni isotopes in radioactive samples fromthe nuclear power plant with anion exchange resin and extraction chromatographic resins. Radionuclides were leached from the radioactive samples withconcentrated HNO3 or HCl. Pu and Am isotopes were sequentially purified withanion exchange resin and TRU resin in HNO3 medium, respectively. Also, Sr and
Tc isotopes were sequentially purified with Sr Resin and TEVA resin in HNO3
medium. Finally, in HCl medium, Nb and Fe isotopes were separated throughanion exchange resin, on the other hand, Ni isotopes were purified with Ni resin.
After α source preparation for the purified Pu and Am isotopes with micro-coprecipitation method, Pu and Am isotopes were measured using alpha spectrometry. Also, Tc, Sr, Fe and Ni isotopes were measured using a low level liquid
scintillation counter, while Nb isotopes were measured by gamma spectrometry.
The radiochemical procedure for Pu, Am, Tc, Sr, Nb, Fe and Ni isotopes investigated in this study has been applied to the radioactive waste samples.

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P8 - 19#8 Radioactive Waste and Current Radiological Issues

Establishment of radioactive waste full cycle history management
system using QR-code

Jin-Woo Lee*, Jong-Jin Kim, Dong-Ju Lee
KAERI (Korea Atomic Energy Research Intitute)1
Email: jinwoo@kaeri.re.kr
Keywords: radioactive waste, full cycle history management system, QR-code,
WCP
The radioactive waste management system using QR code is operated from theproduction of waste in the form of small packaging unit. The Korea Atomic Energy
Research Institute has developed a system that is more effective, and it can bedirectly linked to the field situation with the radioactive waste cycle history management system that was in operation for radioactive waste history management.
The QR code attached on the surface of small packaged radioactive waste cancontain the following information: the year of waste generation, generator, type,
total radioactivity, basic property of contents, and physical characteristics. Many
good effects are expected for a small packaging management system applying
QR code.
In particular, a methodology was established for the steps from the collection
stage to the treatment of radioactive waste so that the radioactive waste certification program (WCP) could be performed. It could be divided into stages suchas waste prediction, waste collection, waste history management, and waste
disposal. For radioactive waste history management could be systematically carried out, it is possible to to reduce the amount of radioactive waste.
This system can be useful more effectively to the acceptance inspection of radioactive waste at the radioactive waste disposal site. And, if information needs
to be checked during the acceptance inspection of radioactive waste at the radioactive waste disposal site, it will be able to be confirmed immediately. Thissystem is beneficial not only to the safety management of radioactive waste generated within KAERI (Korea Atomic Energy Research Intitute) but also to the final
waste disposal of KORAD (Korea radioactive waste agency).

481

Abstracts - Poster

P8 - 20#8 Radioactive Waste and Current Radiological Issues

A Case Study on the Removal and Disposal of RPV from the La
Crosse Boiling Water Reactor Decommissioning Experience.
Jihwan Yu*, Hyung-woo Seo, Gi-lim Kim
Korea Hydro & Nuclear Power (KHNP) Central Research Institute, Republic of Korea1
Email: liu20893283@khnp.co.kr
Keywords: Radiowaste, Decommissioning, RPV(Reactor Presssure Vessel),
BWR(Boiling Water Reactor)
This case study provides detailed information on the decommissioning activities
that have been successfully completed at the La Crosse Boiling Water Reactor(LACBWR), which was shut down in 1987, and the experiences and lessons
learned during decommissioning.
LACBWR was located in southwest Wisconsin on the east shore of the Mississippi
River near the town of Genoa, Wisconsin.
The small 50-Megawatt (MWe) power reactor was first constructed of Wisconsin’s nuclear plant therefore began in May 1963 and completed in 1967. LACBWR
was permanently shut down in 1987, approximately three years before the end of
its operating license, and DPC (Dairyland Power Cooperative) as operator began
to plan for the transition for the SAFSTOR and decommissioning of the plant.
The physical demolition work at LACBWR with the removal of all targeted aboveand below- grade structures was completed by November 2019. The removal ofdemolition debris and soil remediation was also complete by the end of 2019.
As of July 2022, there are some ongoing final reviews of a few Final Status Survey(FSS) report results (some Class 1 survey units) that are being completed by the
NRC.
The study provides the intermittent safe storage (SAFSTOR) and decommissioning process from 1988 through 2014 and the full-scale decommissioning planning and execution activities from 2015 through mid 2022.
The disposition of radiologically contaminated large components presents a
number of logistical handling challenges in removal and disposal based on theirradioactivity, physical size, and weight. The La Crosse Boiling Water Reactor(LACBWR) was a small 50 MWe Boiling Water Reactor (BWR) and its main large
component challenge was the removal of the 150-ton Reactor Pressure Vessel(RPV).
In August 2005, DPC awarded Duratek (a precursor company to Energy Solutions[ES]) a contract for the removal and disposal of the RPV, including the engi-

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P8 - 20

#8 Radioactive Waste and Current Radiological Issues

neering, licensing, plant modifications, and heavy rigging and lifting services toremove the RPV from the Reactor Building and place it on the transport vehicle
for transportation for disposal.
The objective focus was to remove the RPV without segmentation of its internal
components and ship it to the Low-Level Radioactive Waste (LLRW) disposal site
located in Barnwell, SC, in accordance with U.S. Department of Transportation(DOT) criteria.
Key preparation activities for the removal of the RPV included grouting of the RPV,
placement of the temporary lifting device, cutting operations for the Reactor
Building and Bio-shield wall openings, and all associated engineering and logistics planning for the removal of the RPV and the transportation for its disposal.
The internal structure of the RPV was grouted in March 2006. The RPV with thereactor internals, including the 29 control rod blades as well as the forced circulation loop piping and pumps, was filled with Low-Density Cellular Concrete(LDCC) with the reactor head in place. This LDCC fixed in place components andcontamination internal to the RPV. The other plan was to use a Temporary Lifting
Device (TLD). This TLD/Gantry Rail System was erected and installed inside andoutside the Reactor Building.
The runway structure design inside the Reactor Building met the criteria of
NUREG-0612, Control of Heavy Loads at Nuclear Power Plants. Therefore, RPV
removal project activities conducted inside the Reactor Building were performedin compliance with NUREG-0612, due to the close proximity of the RPV lift work
to the stored spent fuel in the spent fuel pool. In compliance with these criteria,
the Reactor Building opening was cut, the reactor shield plug was cut and removal, and a portion of the Bio-shield was cut to prepare the RPV to be moved
out of the Reactor Building. During the RPV removal project, ALARA reviews andengineering measures such as the placement of shields and the use of HEPA
ventilation to reduce external and internal radiation shielding were performed.
Utility personnel did not possess the segmentation or large component removal
expertise. Therefore, specialized contractors were brought in to perform the work.
Recognizing staff expertise limitations, obtaining the rigt guidance and services,and managing a collaborative team enabled DPC to complete this project anddispose of the RPV within the schedule.
It is also noted that recording and obtaining pertinent plant information during
operations supports future decommissioning activities. Careful planning, selection of appropriate tooling, and skilled execution during cutting operations, rigging/lifting, and transport was fundamental in completing this project successfully and safely, without adverse impacts to adjacent spent fuel operations or theenvironment.
Reference
La Crosse Boiling Water Reactor Decommissioning Experience Report: DetailedExperiences 1988-2022. EPRI, Palo Alto, Ca: 2022. 3002023979
483

Abstracts - Poster

P8 - 21#8 Radioactive Waste and Current Radiological Issues

Adsorption Kinetics of Key Nuclides on Geotechnical Media Form
Reprocessing Plant

Tian Xie1
China Institute For Radiation Protection/China1
Email:
Keywords: U,Am-241,Pu-238,Distribution coefficient ,Adsorption kinetics
This study aims to investigate the long-term adsorption kinetics of the key nuclides U, 239Pu, 241Am and 244Cm in environmental media through a series ofadsorption experiments in reprocessing plants, and to analyze the mechanism of
nuclide adsorption by fitting the obtained experimental data with different kinetic
models to provide valuable basic data for the construction of reprocessing plants.
In this experiment, a static batch method was used to perform the experiment, inwhich the adsorbed sample was mixed with a solution of known concentration ofradionuclides, shaken at constant temperature, set different equilibration times,
centrifuged to separate the solid-liquid, and measured the liquid phase nuclide
concentration to obtain the distribution coefficient Kd value. And the adsorption
kinetic process was fitted by quasi-secondary kinetic model and dynamic
boundary model.
The adsorption kinetic process of U; 239Pu; 241Am; 244Cm in fine sand, strongly
weathered granite and silty clay selected in Jiayuguan area, Gansu Province as
representative geotechnical medias were studied by static adsorption experiments. The experimental results were fitted by quasi-secondary order kinetic
model and Moving Boundary Model. The adsorption processes of U; 238Pu;
241Am; 244Cm in three geotechnical medias followed a quasi-secondary orderkinetic model.
The Moving Boundary Model fitting results show that the adsorption of U on finesand and clay media is mainly controlled by liquid film diffusion and the adsorption on strongly weathered granite media is mainly controlled by chemical reactions. The adsorption of 239Pu on fine sand and strongly weathered granite media is mainly controlled by liquid film diffusion, and the adsorption on clay ismainly controlled by chemical reactions. The adsorption of 241Am on three geotechnical media is mainly controlled by liquid film diffusion. The adsorption of
244Cm on three geotechnical media is mainly controlled by intra-particle diffusion.

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#8 Radioactive Waste and Current Radiological Issues

P8 - 22#8 Radioactive Waste and Current Radiological Issues

Study on Rock-Soil Dispersion of Vadose Zone and Aquifer in
Typical Uranium Mining Area in Northern Guangdong

Zhao Ming WANG1, Song YANG1, Wen Jie CHEN1, Chao Chen1,
Ai Ming ZHANG1,*
China Institute for Radiation Protection, China1
Email: 97398574@qq.com
Keywords: uranium mining area, hydrodynamic dispersion, solute penetration
test, H-3
As of 2020, the total reserves of uranium tailings in China have exceeded 60
billion tons, mainly distributed in Guangdong, Hunan, and Jiangxi provinces. Hydrodynamic dispersion is an important mechanism of nuclide migration in rockand soil, and the degree of dispersion is closely related to the type and densityof soil. Accurate dispersion value is the premise of reasonable research on nuclide migration law. In this paper, the dispersion characteristics of nuclides in thesoil and aquifer rock excavated from a typical uranium mining area are studied.
A decommissioned uranium mine in northern Guangdong was selected as an
example to represent a typical uranium mining area, and the undisturbed soilcolumn of sandy clay in the vadose zone and the fully weathered granite in theaquifer were collected on site. The fully weathered granite soil column was filledindoors with the hydraulic conductivity as the main control. The densities, permeability, and soil-water characteristics curves of the two soils were measured;further, 3H was used as a tracer nuclide to study the dynamic nuclide migration.
Using the aerated zone soil column under different rainfall conditions (1, 5, or 10
times to local rainfall). The 3H breakthrough curves of the two soil columns under
different conditions were measured, which were used to obtain the dispersibilityof each test soil column by numerical fitting.
The results show that: in the soil column in the aerated zone, the 3H breakthroughcurve moves to the left with the increase of flow rate, and the curve gradually
becomes steeper. The earlier the soil column reaches the peak concentration,
the higher the peak concentration is. The slower the 3H concentration drops inthe second half of the permeability curve, the more serious the "tailing" phenomenon of the penetration curve is. Under the conditions of 1, 5, and 10 times the
maximum annual rainfall, the hydrodynamic dispersibility DL of the soil column is
1.25, 3.60, 15.00 cm respecticely. The hydrodynamic dispersion of the aquifer

485

Abstracts - Poster

P8 - 22

soil column is 0.77 cm under the condition of groundwater flow velocity in themining area.
Acknowledgement
Financial support for this study was provided by National Key Research and Development Program of the Ministry of Science and Technology of China No.2020Y
FC1806601. This support is gratefully acknowledged.

486

P8 - 23

#8 Radioactive Waste and Current Radiological Issues#8 Radioactive Waste and Current Radiological Issues

Mössbauer Investigation of Low Carbon Steel-Sodium Bentonite
Mixture Aged by γ-Rays Irradiation

Dong LIANG*
Department of Radioactive Waste Management，China Institute for Radiation
Protection，China1
Email: londonliang@yeah.net
Keywords: γ-rays irradiation; low carbon steel-sodium bentonite mixture; mössbauer spectroscopy; high-level radioactive waste；geological disposal
The aim of this paper is to understand the impacts of gamma irradiation on the
corrosion behaviour of X65 low carbon steel in aerobic Gaomiaozi bentonite with
Beishan groundwater and to broaden the assessment and scope of buffer materials for geological disposal of high-level radioactive waste (HLW) in line withthe rapid advancement of HLW repository research.
The bentonite moisture content was adjusted to 17.00% with Beishan groundwater by using the developed freezing mixing method. After at least 3 days ofsealing homogenization, the bentonite powder was mixed with low carbon steel
powder according to the mass ratio of 9:1, 8:2 and 7:3 for use. The samples areevenly loaded into the irradiation container, which can be detected by cleaning
QJ3253-2005 "Test Method for Bubble Leak Detection" after sealing. After passing the test, the irradiation container is put into a large commercial cobalt sourcewith a radiation dose rate of 2.88kGy/h for γ irradiation aging. Take it out after
the accumulated irradiation dose reaches 500kGy, 1000kMGy, 1500kMGy,
2000kMGy, 2500kMGy and 3000kMGy respectively. Disassemble the irradiation
container in the low oxygen glove box, and store the samples sealed immediately
after taking them out. The Mossbauer spectrum of the sample was measured in
iso-accelerated mode, 57Co Mossbauer source based on Pb, and the detector
was proportional to the counter. The test was carried out at room temperature,and the spectral lines were calibrated with α-Fe for velocity, and the isomic shift
took the center of sodium nitroprusside spectrum as the reference point. In orderto ensure the quality of the sampling spectrum, each spectrum was greater than
200,000 counts.MossWinn4.0 was used to fit the spectral lines by least square
method.
The results indicates that, there are four status of Fe in the mixture such as αFe, α-Fe with C in structure, Fe（Ⅱ）and Fe（Ⅲ）. The status of Fe in the mixture is not changed, but the content is varied obviously by γ-rays irradiation aged.

487

Abstracts - Poster

P8 - 23

As a result of oxidation reaction the content of α-Fe have been reduced, and αFe with C in structure is oxidized more thanα-Fe without C in structure; The major
ion of iron is Fe（Ⅲ）, and a number of Fe（Ⅱ）have changed into Fe（Ⅲ）and
there exist transform between Fe- Fe（Ⅱ）- Fe（Ⅲ）possibly.
In general, the content of various Fe in the X65 low carbon steel-GMZ sodium
bentonite (contain groundwater from Beishan area) mixture have been varied after γ-rays irradiation aged, further research is needed to comprehend behaviousof the influence of γ-rays irradiation.

488

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