Role of independent research at AERB for ensuring safety of nuclear facilities in India.

The mission of Atomic Energy Regulatory Board (AERB) of India is to ensure that the use of ionising radiation and nuclear energy in India does not cause unacceptable impact on the workers, members of the public and to the environment. AERB has the mandate to carry out detailed safety review for the siting, construction, commissioning, operation and decommissioning of nuclear and radiation facilities established within the country. To deliver and maintain a strong, credible and technically sound regulation, AERB has established the Safety Research Institute (SRI) at Kalpakkam with a robust technical infrastructure and wide knowledge base. This paper highlights the independent safety research activities carried out at SRI and its role to support and facilitate the decision-making process by AERB at various stages of regulatory review for ensuring safety of the nuclear facilities in India.

Radiation source terms analysis and classification of radioactive waste for the decommissioning of the first HWRR reactor in China.

Nuclear reactors will face the problem of decommissioning at the end of their operating life due to the high radioactivity of reactor components and environmental safety considerations. The Heavy Water Research Reactor (HWRR) is the first large-scale research reactor to be decommissioned in China. The second phase of HWRR decommissioning involves the main components in the reactor block, so the radiation source terms and the radioactive waste level need to be evaluated before demolition and disposal. Based on the operating history, three-dimensional geometry, materials, and other information of the HWRR, the activity of radionuclides in the main components of HWRR is calculated and analyzed, and the MCNP/ORIGEN coupling scheme is utilized for theoretical analysis. The theoretical results indicate that 14C, 54Mn, 55Fe, 60Co, 63Ni, and 152Eu are the main radioactive nuclides. The total activity of radioactive nuclides was 2.36E + 15 Bq at the end of 2007, 4.27E + 13 Bq at the end of 2021, and 1.83E + 13 Bq at the end of 2025. Furthermore, local sampling and radiometric analyses based on the HPGe gamma-ray spectrometer are also performed to verify the theoretical results, the ratio of theoretical activity values to the measured activity of the experimental sample is within 2.5 times, so the theoretical results are conservative. According to the classification standards for radioactive waste, the inner shell, outer shell, cooling water tank, sand layer, and heavy concrete shielding layer are all low-level waste. These results and conclusions can serve as a reference for the second phase decommissioning of the HWRR and the subsequent disposal of radioactive waste.

Study of activation cross sections of double neutron capture reaction on 193Ir for the reactor production route of radiotherapeutic 195mPt.

The radiotherapeutic 195mPt is among the most effective Auger electron emitters of the currently studied radionuclides that have a potential theranostic application in nuclear medicine. Production of 195mPt through double neuron capture of enriched 193Ir followed by ß--decay to the radioisotope of interest carried out at the research reactor IBR-2 is described. Because of the high radiation background, radiochemical purification procedure of 195mPt from bulk of iridium was needed to be developed and is detailed here as well. For the first time, cross section and resonance integral for the reaction 194Ir(n,γ)195mIr were determined. Resonance neutrons contribution was established to exceed that of thermal neutrons, and resonance integral for the reaction 194Ir(n,γ)195mIr is calculated to be 2900 b. Specific activity of 195mPt was estimated to reach a value of 38.7 GBq/(g Pt) at IBR-2 by the end of bombardment (EOB).

High-radiation-exposure tasks in Korean pressurized heavy-water reactors.

This study analysed the occupational dose in Korean pressurized heavy-water reactors (PHWRs) and identified tasks involving high radiation exposure. The average individual dose was sufficiently low to be below the annual effective dose limit for radiation workers and is even lower than the dose limit for the general public. However, some workers received relatively higher doses than others. Furthermore, most PHWR workers are exposed to radiation during planned maintenance periods. In this study, the radiation dose was normalized (radiation dose per unit time) to determine the high-radiation-exposure tasks in Korean PHWRs. Consequently, end-fitting lapping, delayed neutron tube work and fuel channel fixed-end change tasks were identified as high-radiation-exposure tasks in Korean PHWRs. If appropriate radiation protection measures are prioritized for the identified high-dose exposure tasks, optimization of radiological protection will be effectively achieved by reducing the dose that is relatively higher than the average.

Gamma-spectrometry measurements of filtration media from an Advanced Gas-cooled Reactor.

Filtration media used to quantify particulate and gaseous releases have been collected from Hartlepool Power Station in the United Kingdom and measured using high-sensitivity gamma-spectrometry systems. Radionuclides that are relevant to the monitoring regime of the Comprehensive Nuclear-Test-Ban Treaty (CTBT) have been detected. Results are reported and compared to detections recorded on the International Monitoring System (IMS). Time series activity plots have been produced and results interpreted with respect to known plant activities. The reported results improve the understanding of trace-level radionuclide emissions from Advanced Gas-cooled Reactors (AGRs) and aid interpretation of IMS measurements. This work is being performed as part of the Xenon Environmental Nuclide Analysis at Hartlepool (XENAH) collaboration between the Atomic Weapons Establishment (AWE, UK), EDF Energy (UK), Pacific Northwest National Laboratory (PNNL, US) and the Swedish Defence Agency (FOI, Sweden).

United States' regulatory approved pharmacotherapies for nuclear reactor explosions and anthrax-associated bioterrorism.

INTRODUCTION: Nuclear reactor incidents and bioterrorism outbreaks are concerning public health disasters. Little is known about US Food and Drug Administration (FDA)-approved agents that can mitigate consequences of these events. We review FDA data supporting regulatory approvals of these agents. AREAS COVERED: We reviewed pharmaceutical products approved to treat Hematopoietic Acute Radiation Syndrome (H-ARS) and to treat or prevent pulmonary infections following Bacillus anthracis (anthrax) exposure. Four drugs were approved for H-ARS: granulocyte-colony stimulating factor (G-CSF), granulocyte/macrophage colony stimulating factor, pegylated G-CSF, and romiplostim. For bioterrorism-associated anthrax, the FDA approved five antibiotics (doxycycline, penicillin-G, levofloxacin, moxifloxacin, and ciprofloxacin), two monoclonal antibodies (obiltoxaximab and raxibacumab), one polyclonal antitoxin (Anthrax Immune Globulin Intravenous) and two vaccines (Anthrax Vaccine Adsorbed and Anthrax Vaccine Adsorbed with an adjuvant). A national stockpile system ensures that communities have ready access to these agents. Our literature search was based on data included in drugs@FDA (2001-2023). EXPERT OPINION: Two potential mass public health disasters are aerosolized anthrax dissemination and radiological incidents. Five agents authorized for anthrax emergencies only have FDA approval for this indication, five antibiotics have FDA approvals as antibiotics for common infections and for bacillus anthrax, and four agents have regulatory approvals for supportive care for cancer and for radiological incidents.

Progress with the regulation of radiation safety during recovery and removal of spent nuclear fuel from the site for temporary storage at Andreeva Bay on the Kola Peninsula.

In the 1960s, a shore technical base (STB) was established at Andreeva Bay on the Kola Peninsula, in northwest Russia. The STB maintained nuclear submarines and the nuclear icebreaker fleet, receiving and storing fresh and spent nuclear fuel (SNF) as well as solid and liquid radioactive waste (RW). It was subsequently re-designated as a site for temporary storage (STS) for SNF and RW. Over time, the SNF storage facilities partly lost their containment functions, leading to radioactive contamination of workshops and the site above permitted values. The technological and engineering infrastructure at the site was also significantly degraded as well as the condition of the stored SNF. At present, the STS Andreeva Bay facility is under decommissioning. This paper describes progress with the creation of safe working measures for workers involved in site remediation and SNF recovery operations, including the determination of safe shift times in high radiation areas, as part of overall optimization of safety. Results are presented for the successful application of these measures in the period 2019-2021, during which time significant SNF recovery and removal operations were completed without incident. Significant important experience has been gained to support safe removal of remaining SNF, including the most hazardous degraded fuel, as well as recovery of other higher level RW and decommissioning of the old storage buildings and structures.

Production and characterization of 111Ag radioisotope for medical use in a TRIGA Mark II nuclear research reactor.

Radio Pharmaceutical Therapy (RPT) comes forth as a promising technique to treat a wide range of tumors while ensuring low collateral damage to nearby healthy tissues. This kind of cancer therapy exploits the radiation following the decay of a specific radionuclide to deliver a lethal dose to tumor tissues. In the framework of the ISOLPHARM project of INFN, 111Ag was recently proposed as a promising core of a therapeutic radiopharmaceutical. In this paper, the production of 111Ag via neutron activation of 110Pd-enriched samples inside a TRIGA Mark II nuclear research reactor is studied. The radioisotope production is modeled using two different Monte Carlo codes (MCNPX and PHITS) and a stand-alone inventory calculation code FISPACT-II, with different cross section data libraries. The whole process is simulated starting from an MCNP6-based reactor model producing the neutron spectrum and flux in the selected irradiation facility. Moreover, a cost-effective, robust and easy-to-use spectroscopic system, based on a Lanthanum Bromo-Chloride (LBC) inorganic scintillator, is designed and characterized, with the aim of using it, in the future, for the quality control of the ISOLPHARM irradiated targets at the SPES facility of the Legnaro National Laboratories of INFN. natPd and 110Pd-enriched samples are irradiated in the reactor main irradiation facility and spectroscopically characterized using the LBC-based setup and a multiple-fit analysis procedure. Experimental results are compared with theoretical predictions of the developed models, showing that inaccuracies in the available cross section libraries prevent an accurate reproduction of the generated radioisotope activities. Nevertheless, models are normalized to our experimental data allowing for a reliable planning of the 111Ag production in a TRIGA Mark II reactor.

From Nuclear Reactor-Based to Proton Accelerator-Based Therapy: The Finnish Boron Neutron Capture Therapy Experience.

The authors review the results of 249 patients treated with boron neutron capture therapy (BNCT) at the Helsinki University Hospital, Helsinki, Finland, from May 1999 to January 2012 with neutrons obtained from a nuclear reactor source (FiR 1) and using l-boronophenylalanine-fructose (l-BPA-F) as the boron delivery agent. They also describe a new hospital BNCT facility that hosts a proton accelerator-based neutron source for BNCT. Most of the patients treated with nuclear reactor-derived neutrons had either inoperable, locally recurrent head and neck cancer or malignant glioma. In general, l-BPA-F-mediated BNCT was relatively well tolerated with adverse events usually similar to those of conventional radiotherapy. Twenty-eight (96.6%) out of the evaluable 29 patients with head and neck cancer and treated within a clinical trial either responded to BNCT or had tumor growth stabilization for at least 5 months, suggesting efficacy of BNCT in the treatment of this patient population. The new accelerator-based BNCT facility houses a nuBeam neutron source that consists of an electrostatic Cockcroft-Walton-type proton accelerator and a lithium target that converts the proton beam to neutrons. The proton beam energy is 2.6 MeV operating with a current of 30 mA. Treatment planning is based on Monte Carlo simulation and the RayStation treatment planning system. Patient positioning is performed with a 6-axis robotic image-guided system, and in-room imaging is done with a rail-mounted computed tomography scanner. Under normal circumstances, the personnel can enter the treatment room almost immediately after shutting down the proton beam, which improves the unit capacity. ClinicalTrials.gov ID: NCT00114790.

Inadvertent Radiation Exposures in Combat Zones: Risk of Contamination and Radiobiologic Consequences.

On February 24, 2022, Russia began a military invasion of Ukraine. Missile and air strikes were reported throughout the country, shortly followed by a large ground invasion from multiple directions. Four major theaters developed: the Kyiv offensive, the Northeastern Ukraine offensive, the Eastern Ukraine offensive, and the Southern Ukraine offensive, with continued missile and air strikes far into Western Ukraine. Advancing Russian military units launched an attack and captured the Chernobyl nuclear station. Russian troops dug trenches into the area commonly known as the "Red Forest," violating the established radiation safety measures and threatening security within the Chernobyl Exclusion Zone. The placement of military units in such close proximity to the station also sparked concerns of possible damage occurring to the containment vessel constructed around the station's wrecked fourth reactor. There are 15 operating nuclear reactors in Ukraine. Each is vulnerable to an attack or sabotage that could precipitate a malfunction and possible release of radioactive isotopes. In this short commentary, we will discuss radiobiologic data obtained after the analysis of historical nuclear power plant (NPP) accidents and emphasize new challenges for nuclear security when NPPs are found and are possible targets within a conflict zone.

Mortality from various diseases of the circulatory system in the Russian Mayak nuclear worker cohort: 1948-2018.

This paper reports on the findings from the study of mortality from diseases of the circulatory system (DCS) in Russian nuclear workers of the Mayak Production Association (22 377 individuals, 25.4% female) who were hired at the facility between 1948 and 1982 and followed up until the end of 2018. Using the AMFIT module of the EPICURE software, relative risks (RRs) and excess RRs per unit absorbed dose (ERR/Gy) for the entire Mayak cohort, the subcohort of workers who were residents of the dormitory town of Ozyorsk and the subcohort of migrants from Ozyorsk were calculated based on maximum likelihood. The mean cumulative liver absorbed gamma-ray dose from external exposure was 0.45 (0.65) Gy (mean (standard deviation)) for men and 0.37 (0.56) Gy for women. The mean cumulative liver absorbed alpha dose from internal exposure to incorporated plutonium was 0.18 (0.65) Gy for men and 0.40 (1.92) Gy for women. By the end of the follow-up, 6019 deaths with DCS as the main cause of death were registered among Mayak Production Association workers (including 3828 deaths in the subcohort of residents and 2191 deaths in the subcohort of migrants) over 890 132 (622 199/267 933) person-years of follow-up. The linear model that took into account non-radiation factors (sex, attained age, calendar period, smoking status and alcohol drinking status) and alpha radiation dose (via adjusting) did not demonstrate significant associations of mortality from DCS, ischaemic heart disease (IHD) and cerebrovascular disease with gamma-ray exposure dose in the entire cohort, the resident subcohort or the migrant subcohort (either in men or women). For the subcohort of residents, a significant association with gamma dose was observed for mortality from ischaemic stroke in men with ERR/Gy = 0.43 (95% CI 0.08; 0.99); there were no significant associations with liver absorbed gamma dose for any other considered outcomes. As for internal exposure, for men no significant associations of mortality from any DCS with liver absorbed alpha dose were observed, but for women positive associations were found for mortality from DCS (the entire cohort and the resident subcohort) and IHD (the entire cohort). No significant associations of mortality from various types of DCS with neutron dose were observed either in men or women, although neutron absorbed doses were recorded in only 18% of the workers.

Beam port filters in a TRIGA MARK III nuclear reactor to produce epithermal neutrons for BNCT.

Glioblastoma multiforme is the most common and aggressive brain tumor and it is difficult to treat with conventional surgery, chemotherapy, or radiation therapy. An alternative treatment is boron neutron capture therapy which requires an energy modulated beam of neutrons and a10B drug capable of adhering to the tumor. In this work, MCNP6 Monte Carlo code was used to evaluate the effect on the neutron spectrum by placing two filters along the radial beam tube of the TRIGA Mark III nuclear reactor of ININ in Mexico. Every filter was made with the same amount and type of materials: Steel and Graphite for filter 1 and Cadmium, Aluminum, and Cadmium (Cd + Al + Cd) for filter 2. Two cases were analyzed for each filter as follows: Case A for filter 1 was considering 30 cm of steel and 30 cm of graphite, while for case B, the dimensions of filter 1 were 15 cm of steel, 15 cm of graphite, 15 cm of steel and 15 cm of graphite. Cases A and B for filter 2 were analyzed considering the same dimensions and amount of materials. The work was in the aim to produce epithermal neutrons for boron neutron capture therapy. Neutron spectra were calculated at three sites along the beam tube and two sites outside the beam tube; here, the ambient dose equivalent, the personal dose equivalent, and the effective doses were also estimated. At a distance of 517 cm of core, in case B, results in an epithermal-to-thermal neutron fluence ratio of 30.39 was obtained being larger than the one recommended by the IAEA of 20.

Performance of radio-iodine discharge control methods of nuclear reprocessing plants.

It is imperative to control radio-iodine discharges to atmosphere from nuclear reprocessing plants. Inhalation and ingestion of radio-iodine cause its concentration in the thyroid gland leading to risk of thyroid cancer in humans. Two isotopes of iodine viz. iodine-131 (131I) and iodine-129 (129I) are generated in considerable quantities in the nuclear fuel as fission products in the nuclear reactors. From nuclear reactors, no iodine is released to the atmosphere during normal operations, whereas from spent fuel reprocessing plants, during normal operation, iodine is discharged to the atmosphere, mainly through gaseous discharges. Shortly after the initial periods of reprocessing in 1944, iodine emission control methods were incorporated in the design of reprocessing plants. At the time of spent fuel discharge from reactor, quantity of 131I is high and can contribute radiation dose to humans during reprocessing operations. A delay or cooling period of spent fuel, before reprocessing for a definite number of days can reduce the quantities to below the permissible limits of discharge due to its short half-life of 8 days. 129I has a very long half-life, and is only significant for reprocessing plants of large throughput and high fuel burn-ups. Minimum required de-contamination factor (DF) for iodine for a reprocessing plant can be estimated from the limits of discharge of iodine stipulated by regulatory authority of each country. Though many processes were developed and demonstrated extensively in lab and pilot scale, only a few of these processes were found to be suitable for commercial deployment. This paper reviews systematically the operation experiences and performance characteristics of iodine control methods implemented so far. The review also focus on the effect of integrating various iodine control methods on the main reprocessing operations and thereby facilitate selection of the optimum iodine control method.

Brain cancer incidence rates and the presence of nuclear reactors in US states: a hypothesis-generating study.

BACKGROUND: The etiology of brain cancer is poorly understood. The only confirmed environmental risk factor is exposure to ionizing radiation. Because nuclear reactors emit ionizing radiation, we examined brain cancer incidence rates in the USA in relation to the presence of nuclear reactors per state. METHODS: Data on brain cancer incidence rates per state for Whites by sex for three age groups (all ages, 50 and older, and under 50) were obtained from cancer registries. The location, number, and type of nuclear reactor, i.e., power or research reactor, was obtained from public sources. We examined the association between these variables using multivariate linear regression and ANOVA. RESULTS: Brain cancer incidence rates were not associated with the number of nuclear power reactors. Conversely, incidence rates per state increased with the number of nuclear research reactors. This was significant for both sexes combined and for males in the 'all ages' category (ß = 0.08, p = 0.0319 and ß = 0.12, p = 0.0277, respectively), and for both sexes combined in the'50 and older' category (ß = 0.18, p = 0.0163). Brain cancer incidence rates for counties with research reactors were significantly higher than the corresponding rates for their states overall (p = 0.0140). These findings were not explicable by known confounders. CONCLUSIONS: Brain cancer incidence rates are positively associated with the number of nuclear research reactors per state. These findings merit further exploration and suggest new opportunities for research in brain cancer epidemiology.

Nuclear data for reactor production of 131Ba and 133Ba.

The newest radioisotope for brachytherapy treatment of prostate cancer is 131Cs (t1/2 = 9.69 d, 100% EC). Generated via electron capture decay of 131Ba (t1/2 = 11.6 d, 100% EC), 131Cs has been used in brachytherapy for prostate cancer since 2004. The 131Ba parent is produced through neutron capture of enriched 130Ba in a nuclear reactor. For large-scale production of 131Ba, an accurate knowledge of production and burnup cross sections of 131Ba are essential. In this paper, we report two group cross sections (thermal and resonance integrals) for 130Ba and 131Ba and a new measure of the half-life of 131Ba. Targets consisting of milligram quantities of enriched 130Ba (∼35%) were irradiated in Oak Ridge National Laboratory's High Flux Isotope Reactor at thermal and resonance neutron fluxes of (1.9-2.1) × 1015 and (5.8-7.0) × 1013 neutrons·cm-2 s-1, respectively, for durations ranging from 3 to 26 days. In addition, cadmium covered samples of 130Ba were irradiated for 1 hour at 12.6% full reactor power (10.7 MW). The yield of 131Ba approaches a saturation value of ∼60 GBq (∼1.6 Ci) per mg of 130Ba for 20 days irradiation at a thermal neutron flux of 1.8 × 1015 n·s-1·cm-2, with a thermal/epithermal ratio of ∼30. Under the above experimental conditions, the two group cross sections of 130Ba are 6.9 ±â€¯0.5 b (thermal, σ0) and 173 ±â€¯7 b (resonance, I0). These values represent the sum of cross sections to metastable and ground states of 131Ba. For 131Ba, the empirically measured thermal cross section is 200 ±â€¯50 b assuming an I0/σ0 of 10. This cross section is reported for the first time. Further, the half-life of 131Ba was remeasured to be 11.657 ±â€¯0.008 d. Lastly, this study also resulted in the co-production of 133Ba (t1/2 = 10.52 y, 100% EC). The experimental yield of 133Ba is ∼370 MBq (∼10 mCi) per mg of 132Ba (thin target) for one cycle irradiation in the High Flux Isotope Reactor, and measured two-group 132Ba cross sections are 7.2 ±â€¯0.2 b and 39.9 ±â€¯1.3 b. These values also represent the sum of cross sections to metastable and ground states of 133Ba.

Lung Cancer in the Mayak Workers Cohort: Risk Estimation and Uncertainty Analysis.

The workers at the Mayak nuclear facility near Ozyorsk, Russia are a primary source of information about exposure to radiation at low-dose rates, since they were subject to protracted exposures to external gamma rays and to internal exposures from plutonium inhalation. Here we re-examine lung cancer mortality rates and assess the effects of external gamma and internal plutonium exposures using recently developed Monte Carlo dosimetry systems. Using individual lagged mean annual lung doses computed from the dose realizations, we fit excess relative risk (ERR) models to the lung cancer mortality data for the Mayak Workers Cohort using risk-modeling software. We then used the corrected-information matrix (CIM) approach to widen the confidence intervals of ERR by taking into account the uncertainty in doses represented by multiple realizations from the Monte Carlo dosimetry systems. Findings of this work revealed that there were 930 lung cancer deaths during follow-up. Plutonium lung doses (but not gamma doses) were generally higher in the new dosimetry systems than those used in the previous analysis. This led to a reduction in the risk per unit dose compared to prior estimates. The estimated ERR/Gy for external gamma-ray exposure was 0.19 (95% CI: 0.07 to 0.31) for both sexes combined, while the ERR/Gy for internal exposures based on mean plutonium doses were 3.5 (95% CI: 2.3 to 4.6) and 8.9 (95% CI: 3.4 to 14) for males and females at attained age 60. Accounting for uncertainty in dose had little effect on the confidence intervals for the ERR associated with gamma-ray exposure, but had a marked impact on confidence intervals, particularly the upper bounds, for the effect of plutonium exposure [adjusted 95% CIs: 1.5 to 8.9 for males and 2.7 to 28 for females]. In conclusion, lung cancer rates increased significantly with both external gamma-ray and internal plutonium exposures. Accounting for the effects of dose uncertainty markedly increased the width of the confidence intervals for the plutonium dose response but had little impact on the external gamma dose effect estimate. Adjusting risk estimate confidence intervals using CIM provides a solution to the important problem of dose uncertainty. This work demonstrates, for the first time, that it is possible and practical to use our recently developed CIM method to make such adjustments in a large cohort study.

Database of Families of Workers Chronically Exposed to Radiation: Data and Biospecimen Resources.

ABSTRACT: Animal experiment findings suggest that high doses of ionizing radiation exposure (>1.0 Gy) may cause genetic and epigenetic effects in offspring. However, epidemiological studies of offspring of radiation-exposed parents did not find increased risks of any health effects. Findings of cellular/experimental investigations and studies of human health effects are contradicting, and further investigations are needed to help resolve ambiguities using updated and/or improved data. This paper provides a detailed description of a database of families of workers of the first Russian nuclear facility, Mayak Production Association, located in the Southern Urals in the Chelyabinsk region close to Ozyorsk city, which started its operation in 1948 and today consists of reactors, radiochemical and plutonium production plants, and auxiliary facilities. The Mayak worker cohort includes 22,377 individuals (25% females) who were hired at one of the main Mayak PA facilities between 1948 and 1982 and were externally or internally exposed to ionizing radiation over prolonged periods. Advantages of the cohort include its large size, extensive follow-up period (70 y), individually measured doses from external and internal exposure and the wide range of these doses, heterogeneity by gender/age/ethnicity/initial health status, complete data on vital status and causes of death, available medical information on morbidity and reproduction, available data on non-radiation factors, and stored biological specimens donated by more than one-third of the cohort members. Based on medical and dosimetry database "Clinics" containing raw data on workers of the study cohort, the Mayak workers' family and offspring database was created. To date, it comprises 12,195 family couples (a husband and a wife) and 16,585 offspring. Biological specimens are available for more than 1,000 family triads (a husband, a wife, and their child). Stages of assembling the database and its descriptive characteristics are presented in this paper. Examples of potential applications of the database for investigations of non-targeted and transgenerational radiation effects in offspring of exposed parents are discussed.

ESTIMATION OF PHOTON ENERGY AND DIRECTION DISTRIBUTIONS AT JAPANESE NUCLEAR POWER PLANTS BASED ON LITERATURE SURVEY FOR J-EPISODE STUDY.

In order to reconstruct organ-absorbed dose from recorded dose for risk estimation in nuclear worker cohort, the preceding study of the International Agency for Research on Cancer (IARC) 15-Country Collaborative Study estimated the organ dose conversion factor from the recorded dose of Hp(10) under the assumption that on average, in the nuclear power plants (NPPs), 10% of the dose received by workers was due to photon energies ranging from 100 to 300 keV and 90% from photon energies ranging from 300 to 3000 keV, with the average geometry being 50% in the antero-posterior geometry and 50% in the isotropic geometry. Similar examination was conducted at the Japanese Epidemiological Study on Low-Dose Radiation Effects (J-EPISODE). Literature survey disclosed that Japanese electric power companies had jointly conducted the research on energy distribution and incidence direction distribution of gamma rays in working environments during periodical inspection and maintenance as well as during operation in the 1980s. The analysis of the survey results on photon energy and geometry distribution of Japanese NPPs demonstrated appropriateness in applying the IARC study assumption for nuclear workers in Japan and reconstructing organ-absorbed dose in the J-EPISODE. These results in Japan also provide strong evidence to support the robustness and generality of the IARC study assumption, which was estimated based on the judgment of experts at nuclear facilities around the world.

The overview and prospects of BNCT facility at Tsing Hua Open-pool reactor.

The whole picture of the BNCT facility at Tsing Hua Open-pool Reactor will be presented which consists of the following aspects: the construction project, the beam quality, routine operations including the QA program for the beam delivery, determination of boron-10 concentration in blood, T/N ratio, and the clinical affairs including the patient recruit procedure and the patient irradiation procedure. The facility is positioned to serve for conducting clinical trials, emergent (compassionate) treatments, and R&D works.