Vol.25,

The absolute dose difference of the iso-center position at different times was used to evaluate the stability of medical linear accelerator (LINAC). In order to eliminate the drawback of point dose without statistical significance and to investigate the output stability of segment with small weight related to intensity-modulated radiotherapy (IMRT), a modified method of evaluating the stability of LINAC was proposed based on planar dose combined with gamma method. With a commercial available two-dimensional ionization chamber, a set of planar doses with dose gradient from 2 cGy to 100 cGy were obtained. Then the gamma method was adopted to analyze the dose difference between the reference data and the evaluated data at each dose level. The results showed that the improved method based on planar dose for analyzing the output stability of the medical LINAC was feasible and efficient, and suggested that the reverse optimization should be aborted in clinical when the segment weight related to IMRT was under 10 MU.

456

Induced radioactivity is one of the essential problems in the radiation protection field of proton accelerators. Research on the induced radioactivity of low-energy proton accelerators is highly limited. Given such context, this study investigates the cross section of ⁶³Cu (p, n) ⁶³Zn and ⁶⁵Cu (p, n) ⁶⁵Zn in Cu targets at 11 MeV proton accelerators through an activation experiment. The uncertainties of the results are analyzed in detail. Results show that the cross section of the reaction of ⁶⁵Cu (p, n) ⁶⁵Zn in the experiment is only 1.36% lower than that of the FLUKA simulation, whereas the reaction of ⁶³Cu (p, n) ⁶³Zn in the experiment is 25.4% higher than that of the FLUKA simulation. Given that the benchmark for the FLUKA code of low-energy proton accelerators is very limited, this study provides a reference in this field.

440

Since the development of positron emission tomography in 1937, the number of medical cyclotrons used in producing radioisotopes (RI) has risen sharply. In its operation, the structural materials of a medical cyclotron and the concrete shielding of its vault are activated by neutrons. In analyzing domestic and international decommissioning practices, the dismantling characteristics of medical cyclotrons result in a large amount of radioactive wastes, such as concrete, although their activity level is very low. In the United States, the dismantling plan must be established by the applicant. The decommissioning funding plan (DFP) is a financial assurance demonstration based on a site-specific cost estimate of decommissioning the facility. In 10 CFR 30.35, NRC requires the submission of DFP in the case of the possession and use of unsealed byproduct materials with a half-life of greater than 120 days and in quantities exceeding 10⁵ times that of the applicable quantities. The dismantling of the Seoul National University Hospital (SNUH) cyclotron (TR-13) was performed in December 2012. Some of the risks resulting from this dismantling are related to radioactive wastes, as well as issues concerning dismantling costs. Hence, we propose introducing the DFP at the authorization stage. DFP will be helpful in preparing reliable decommissioning plans for safe decommissioning and unexpected early decommissioning in the future. DFP can also contribute in reducing radioactive waste and in decreasing the decommissioning cost by preventing excessive establishments. This study provided an overview of the decommissioning aspects of the SNUH cyclotron and the necessity of adopting the DFP for decommissioning medical cyclotrons.

505

For recent decades, a considerable amount of ^99mTc has been stimulated for diagnosis radiopharmaceuticals because of its physical advantages. The increase in the use of ^99mTc makes naturally more chances of internal exposure for not only patients but also medical workers. The patient internal exposure by an intravenous injection has been assessed relatively well with the reports of International Commission on Radiological Protection (ICRP) or Medical Internal Radiation Dose(MIRD). However, there are few studies which can support its accurate assessment for medical worker who treats ^99mTc. In spite of the absence of information, the physiological information of each ^99mTc-labelled compound for patient, provided by ICRP, can be used optionally for worker exposure because the behaviors after uptake to blood are similar for patient and worker. Using the data, in this study, the data for bioassay were given as the intake whole-body retention and urinary excretion function. We selected the two most frequently used ^99mTc-labelled compounds based on statistical data; ^99mTc-phosphonate, pertechnetate. The data of the Human Alimentary Tract Model (HATM, publication 100) and the revised Human Respiratory Tract Model (revised HRTM, OIR) were used for compartment models which depict the physiological behavior in body after intake. In case of ^99mTc-phosphonate, we adopted the systemic model for patient intake described in ICRP publication 53 based on the assumption that the behaviors after uptake to blood are similar for patient and worker. On the other hand, recent updated systemic model in OIR report could be directly adopted for pertechnetate. We used Birchall’s algorithm for calculation and developed the module which could calculate the retention amounts in each compartment at given time by MATLAB. In addition, we fitted the functions as sum of exponential terms using ORIGIN and the fitting coefficients were provided. We also could calculate the committed dose coefficients for each compound using SAF values for photons provided by Cristy and Eckerman. The results in this study will be useful to estimate the intake and effective dose for medical field.

371

A monitoring system based on a CaF₂ flow-cell detector was previously developed for measuring the tritium concentration in water. The flow-cell detector was fabricated using a granular CaF₂(Eu) solid scintillator. In the present study, the effect of lead shielding on the measurement sensitivity was investigated. First, measurements were carried out both on tritium water and tritium-free water without the use of shielding. Counts were accumulated for either 600 or 10 000 s and the detection limits without shielding were calculated. Second, the effect of shielding was evaluated using background radiation. Lead blocks with a thickness of 5 cm were used as shielding materials. Full shielding around the entire measurement system reduced the background count by 53.8%. Approximating this to a background reduction of 2 times, the detection limit with full shielding was then calculated. It was concluded that 5-cm-thick lead shielding reduced the detection limit for tritium by at least 35%

499

A concept of space radiation-voltaic batteries (SRVBs) based on various space radiation environments, including the earth radiation belt ( ERB), solar proton event ( SPE), and galactic cosmic rays ( GCRs) is presented in this paper. The energy deposition, structural parameter optimization, and output performance of SRVBs are studied through the Monte Carlo toolkit Geant4 and theoretical formulas. Energy deposition in the three radiation environments monotonically decreases as the depth in both silicon film models and energy conversion units of the SRVBs increase. The recommended junction depths of SRVBs based on ERB, SPE, and GCR are 0.5, 3.0, and 0.5 μm, respectively. In the proposed design scheme, the maximum output power density of SRVBs in ERB, SPE, and GCR are 2.022×10^-7, 1.113×10^-3, and 3.844×10^-17 W⋅cm^-2, respectively. Calculation results suggest that SRVBs based on ERB and SPE are high-potential candidates for space power sources. The results could help raise awareness on the use of space radiation particle energy and facilitate research on SRVBs.

333

The Monte Carlo method can be widely applied to particle transport through numerous simulated data processing operations. However, this process consumes much time. Traditional parallel computing based on multi-CPU or multi-core CPU can effectively address this issue, but it is limited by inadequate computer hardware. Nonetheless, the current programmability and parallel processing capability of digital graphics processing units (GPUs) can sustain general computing applications such as Monte Carlo program simulation. This paper presents a method that facilitates the parallel computation of the Monte Carlo procedure through GPUs. Its feasibility is verified through a sample of simplified photon transport program, the results indicate that execution time can be shortened by approximately 90 times. Based on the general Monte Carlo program Geant4, the photon and electronic coupled transport module was examined, analyzed, and rewritten using the GPU programming language OpenCL to generate a Geant4 parallel tool [base on GPU parallel computing tool (BOGPT)]. The simulation results of the standard examples demonstrated that the outcomes of the BOGPT program are similar to those of Geant4 and the simulation time can be reduced by approximately three times. Finally, the GPU programming-based parallel computing method for Monte Carlo applications is accelerated and implementation prospects are broadened following further optimization.

543

The positron annihilation lifetime (PAL) technique has been applied to study the effects of gamma-ray irradiation on polymer materials, such as ethylene propylene diene monomer (EPDM), nitrile butadiene rubber (NBR), and chloroprene rubber (CR), used in various industries. PAL measurements were performed with a conventional fast-fast coincidence system. Polymer materials were irradiated with 25 Gy/h in three conditions (50, 100, 500 Gy) and 5 kGy/h in six conditions (5, 50, 200, 500, 800, and 1 000 kGy) at room temperature with a ⁶⁰Co gamma-ray source. Afterwards, the irradiated samples were measured by PAL spectroscopy. The delayed lifetime spectra of irradiated and unirradiated samples were decomposed into three lifetime components. The lifetime parameters indicate the free-volume size and concentration. The longest lifetime component represents the free volume size in polymer. Free volume and intensity were multiplied as a measure of relative fraction of free-volume hole. Cross-linking and degradation could occur when a polymer is irradiated. The free volumes of polymers were decreased accordingly. These results were verified through PAL measurement. From the Tao-Eldrup model, the mean radius and free volume fraction of CR, EPDM, and NBR were calculated. We certified the decrease of the fractional free volume. From these results, this study verified the feasibility of PA application on polymers.

369

In the process of neutron spectrum measurement using the multi-sphere neutron spectrometer, energy response functions and detector readings should be applied to neutron spectrum unfolding. Mathematically, there can be multiple solutions to this problem, but only one actual neutron spectrum exists. Compared with common numerical spectrum unfolding methods, genetic algorithms have the characteristics of global optimization and probabilistic search. Therefore, they are chosen to be the spectrum unfolding algorithms for the multi-sphere neutron spectrometer (MNS IL100) developed by Tsinghua University. Firstly, the detector and different size polyethylene spheres of MNS IL100 were modeled to calculate the energy response functions by applying Monte Carlo simulation. Then based on the physical and mathematical properties of the spectrum unfolding problem by using genetic algorithms, effective search space and proper fitness function were determined to improve the efficiency of search and iteration. The elitism replacement scheme was used to ensure convergence and the pseudo-parallel strategy was used to inhibit premature convergence. According to the algorithms mentioned above, a spectrum unfolding code was developed and tested with several typical neutron spectra. At last, MNS IL100 and the spectrum unfolding code were used in actual experiment of ²⁵²Cf neutron source spectrum measurement. The experimental result is in good agreement with the ²⁵²Cf standard neutron spectrum, verifying the effectiveness and practicality of using genetic algorithms to unfold the neutron spectrum with combination of several processing strategies.

388

The radiation shielding calculations of large nuclear facilities are complicated due to their bulk shields and complex geometries. A program system based on three-dimensional discrete ordinates (S_N)-Monte Carlo (MC) coupled method has been developed to solve this kind of shielding problems. S_N method is used to treat the deep penetration problem in the bulk shield, and MC simulation is performed with complex geometry. The coupled method is implemented in the interface program to calculate the probability of particle variables from the S_N angular flux distribution. A special source routine is written and linked to MC code to processed these probability arrays and sample the particle variables to generate source for the use of MC code. The H. B. Robinson Unit 2 pressure vessel benchmark was used to validate the S_N-MC coupled program system. The specific activities for the six dosimeters in the surveillance capsule were calculated. Satisfactory agreements were obtained from the comparisons of S_N-MC results with those obtained from measurement and other computer codes. The result demonstrates that the S_N-MC coupling scheme with the program system is suitable to treat three dimensional shielding problems with satisfactory accuracy.

549

This paper presents the mitigation effects of fission product reduction strategies that can be applied to severe accident management for CANDU6 plants. Strategy effects for 'control of the reactor building condition’ and 'reduction of the fission product release’ have been evaluated. The operations of a dousing spray system, local air cooler system, and containment filtered vent system are simulated as severe accident management strategies. The present analysis is mainly focused on determining an estimate of the fission products in the release and transport processes, i.e., (1) their release from the reactor core, (2) their distribution in a primary heat transport system, calandria vessel, molten corium, and reactor building, and (3) their release from the reactor building into the environment during a severe accident.

321

After the Fukushima nuclear accident in 2011, improvement actions of the environmental radiation surveillance were carried out by Chinese government to deal with the possible nuclear accident in response to the rapid development of nuclear power in China. The local government radiation surveillance system, including the on-line radiation monitoring network and automatic sampling system, the off-site monitoring center laboratory, the radioactive effluent on-line monitoring system and the sampling inspection laboratory were regulated to establish for all operation and constructing nuclear power plant. This paper describes the general design of the system by taking Ningde nuclear power plant (NPP) for example. The main designs, including radiation monitoring and sampling equipment, data collection and the communication technology, and the surveillance management, are generally based on the experiences or lessons from Fukushima accident. The system is expected to act as a pivotal role to evaluate the environmental radioactivity from the operation of NPP, and to provide effective decision support in the event of possible nuclear accident.

389

In recent years, neutrons have been studied for application in fields such as material analysis and boron neutron capture therapy. To create a compact shield for these facilities, a neutron shield concrete is developed. Verifying the homogeneity of the concrete is important to ensure adequate shielding performance. In this research, neutron radiography images of the concrete are taken using the Thermal Neutron Radiography Facility (TNRF) of the JRR-3 research reactor, and the transmission ratio of the thermal neutrons were estimated. The results showed that the transmission ratio of the concrete was almost the same at each depth.

361

Radiation shielding structure for C-ADS Inject-I was designed and optimized using FLUKA code. For this equipment was planned to be established in an existing tunnel hall and installed in limited space, several shielding hot spots are re-designed and analyzed respectively as they may cause radiation dose leakage and weakening of the total shielding effect. Besides, some new shielding structures are applied in the simulation process. All designed shielding structures are discussed in this paper and the results will meet the governmental criteria for radiation protection.

388

In order to monitor the diffusion and deposition of radioactive plume in Qinshan area after Fukushima nuclear accident, a continuous air sampling was did from March 25 to May 6. A trace of ¹³⁴Cs and ¹³¹I were detected in the period. Results of conventional environmental radiation monitoring also showed that the specific activity of ⁹⁰Sr and ¹³⁷Cs in environmental samples had some abnormal changes in 2011 which almost felt back to normal level in 2012. Although there is no abnormal change on environmental radiation level of Qinshan area till now, the influence caused by Fukushima nuclear accident should be paid continuously attention in future by radiation monitoring.

483