• Nuclear Engineering and Technology
• /
• v.56 no.3
• /
• pp.1029-1036
• /
• 2024

Drawing on the deep experience and understanding of the principles of nuclear safety, as well as many years of nuclear power plant design and operation, the EDF led NUWARD SMR Project is developing a design for a Small Modular Reactor (SMR) of 340 MWe composed of two 170 MWe independent units, that will supplement the offering of high-output nuclear reactors, especially in response to specific needs such as replacement of fossil-fuelled power plants. NUWARD SMR is a mix of proven and innovative design features that will make it more commercially competitive, while integrating safety features that comply with the highest international standards. Following the principles of redundancy and diversity and rigorous application of Defence in Depth (DID), with an international view on nuclear safety licensing, the Project also incorporates new safety approaches into its design development. The NUWARD SMR Project has been in development for a number of years, it entered conceptual design formally in mid-2019 and entered Basic Design in 2023. The objective of the concept design phase was to confirm the project technological choices and to define the first design configuration of the NUWARD SMR product, to document it, in order to launch pre-licensing with the French Safety Authority (ASN) and to define its estimated cost and its subsequent development and construction schedules. As a delivery milestone the Safety Options file (called the Dossier d'Options de Sûreté (DOS)) has been submitted to ASN in July 2023 for their opinion. An integral part of the NUWARD SMR Project, is not only to deliver a design suitable for France and to satisfy French regulation, but to develop a product suitable and indeed desirable, for the international market, with a first focus in Europe. In order to achieve its objectives and realise its market potential, the NUWARD SMR Project needs to define and realise its safety approach within an international environment and that is the key subject of this paper. The following paper: • Summarises the foundation principles and technological background which underpin the design; • Contextualises the key design features with regard to the international safety regulatory framework with particular emphasis on innovative passive safety aspects; • Illustrates the Project activities in preparation for first licensing in France, and also a wider international view via the ASN led Joint Early Review of the NUWARD SMR design, including Finnish and Czech Republic regulators, recently joined by the Swedish, Polish and Dutch regulators; • Articulates the collaborative approach to design development from involvement with the Project partners (the Commissariat à l'énergie atomique et aux énergies alternatives (CEA), Naval Group, TechnicAtome, Framatome and Tractebel) to the establishment of the International NUWARD Advisory Board (INAB), to gain greater international insight and advice; • Concludes with the focus on next steps into detailed design development, standardisation of the design and its simplification to enhance its commercial competitiveness in a context of further harmonisation of the nuclear safety and licensing requirements and aspirations.

• Journal of the Korean Society of Radiology
• /
• v.16 no.3
• /
• pp.327-334
• /
• 2022

In this study, the purpose of this study was to conduct a basic study on the effectiveness and feasibility of the regulation of the Nuclear Safety Act for the department of radiology by examining the questionnaire on the satisfaction of on-campus practice while attending the department of radiology and the current status of radiation workers and radiation related workers. As for the satisfaction of the workers who were designated as frequent visitors while attending the department of radiology and did not handle and operate the radiation generator during on-campus training, 34.62% of the workers answered 'not satisfied'. On the other hand, 50% of workers who were designated as radiation workers while attending school or who were enrolled in school before the regulation of the nuclear safety act and handled and operated radiation generators were 'satisfied' at 50%. In addition, the annual exposure dose of radiation workers in educational institutions was found to be less than 0.05 mSv. If you look at the trends of radiation workers and radiation workers, it can be seen that students who graduate from the Department of Radiology find the most employment in the field dealing with diagnostic radiation generators registered as radiation workers among medical institutions. Therefore, by easing the regulations of the current Nuclear Safety Act or by amending the medical act and the rules on the safety management of diagnostic radiation generating devices, etc. It is presumed that something is necessary.

• Nuclear Engineering and Technology
• /
• v.45 no.5
• /
• pp.573-580
• /
• 2013

The combined effects of reactivity coefficients, along with other core nuclear characteristics, determine reactor core behavior in normal operation and accident conditions. The Power Coefficient of Reactivity (PCR) is an aggregate indicator representing the change in reactor core reactivity per unit change in reactor power. It is an integral quantity which captures the contributions of the fuel temperature, coolant void, and coolant temperature reactivity feedbacks. All nuclear reactor designs provide a balance between their inherent nuclear characteristics and the engineered reactivity control features, to ensure that changes in reactivity under all operating conditions are maintained within a safe range. The $CANDU^{(R)}$ reactor design takes advantage of its inherent nuclear characteristics, namely a small magnitude of reactivity coefficients, minimal excess reactivity, and very long prompt neutron lifetime, to mitigate the demand on the engineered systems for controlling reactivity and responding to accidents. In particular, CANDU reactors have always taken advantage of the small value of the PCR associated with their design characteristics, such that the overall design and safety characteristics of the reactor are not sensitive to the value of the PCR. For other reactor design concepts a PCR which is both large and negative is an important aspect in the design of their engineered systems for controlling reactivity. It will be demonstrated that during Loss of Regulation Control (LORC) and Large Break Loss of Coolant Accident (LBLOCA) events, the impact of variations in power coefficient, including a hypothesized larger than estimated PCR, has no safety-significance for CANDU reactor design. Since the CANDU 6 PCR is small, variations in the range of values for PCR on the performance or safety of the reactor are not significant.

• The Korean Journal of Nuclear Medicine Technology
• /
• v.18 no.2
• /
• pp.57-61
• /
• 2014

Purpose Iodine (I-131) is one of the most widely used radioactive isotopes for therapeutic in the field of nuclear medicine. Therapeutic I-131 capsule is made out of lead to shield high energy radiation. Accurate dosimetry is necessarily required to perform safe and effective work for relative workers. The Monte Carlo method is known as a method to predict the absorbed dose distribution most accurately in radiation therapy and many researchers constantly attempt to apply this method to the dose calculation of radiotherapy recently. This paper aims to calculate distance dependent and activity dependent therapeutic I-131 capsule using GEANT4. Materials and Methods Therapeutic capsules was implemented on the basis of the design drawings. The simulated dose was determined by generating of gamma rays of energy to more than 364 keV. The simulated dose from the capsule at the distance of 10 cm and 100 cm was measured and calculated in the model of water phantom. The simulated dose were separately calculated for each position of each detector. Results According to the domestic regulation on radiation safety, the dose at 10 cm and 100 cm away from the surface of therapeutic I-131 capsule should not exceed 2.0 mSv/h and 0.02 mSv/h, respectively. The simulated doses turned out to be less than the limit, satisfying the domestic regulation. Conclusion These simulation results may serve as useful data in the prediction of hands dose absorbed by I-131 capsule handling. GEANT4 is considered that it will be effectively used in order to check the radiation dose.

• Nuclear Engineering and Technology
• /
• v.55 no.12
• /
• pp.4367-4381
• /
• 2023

Nuclear power plants are increasingly being equipped with digital I&C systems. Although some probabilistic safety assessment (PSA) models for the digital I&C of nuclear power plants have been constructed, there is currently no specific internationally agreed guidance for their modeling. This paper presents an initiative by the OECD Nuclear Energy Agency called "Digital I&C PSA - Comparative application of DIGital I&C Modelling Approaches for PSA (DIGMAP)", which aimed to advance the field towards practical and defendable modeling principles. The task, carried out in 2017-2021, used a simplified description of a plant focusing on the digital I&C systems important to safety, for which the participating organizations independently developed their own PSA models. Through comparison of the PSA models, sensitivity analyses as well as observations throughout the whole activity, both qualitative and quantitative lessons were learned. These include insights on failure behavior of digital I&C systems, experience from models with different levels of abstraction, benefits from benchmarking as well as major contributors to the core damage frequency and those with minor effect. The study also highlighted the challenges with modeling of large common cause component groups and the difficulties associated with estimation of key software and common cause failure parameters.

• Journal of the Korean Society of Radiology
• /
• v.15 no.3
• /
• pp.355-359
• /
• 2021

In this study, the purpose of this study was to analyze the degree of exposure of radiation workers assigned to the Department of Radiology and frequent visitors during on-campus practice, and to conduct a basic study on the feasibility and optimization of the radiation protection of the Nuclear Safety Act for the Department of Radiology. . The average exposure dose of occupational workers by year was 0.01 mSv, the lowest in 2014 and 2016. The highest figure was 0.12 mSv in 2018. The average exposure dose of frequent visitors by year was the lowest at 0.013 mSv in 2018, and the highest at 0.022 mSv in 2016. According to this study, the annual exposure dose received by professors, practical assistants, and students in the department of radiology (department) who use only radiation generators in the course of in-school practice is less than 1 mSv, which is the dose limit for the general public. Therefore, at the time when the radiation dose of students in the Department of Radiology is lower than the dose limit of the general public, the current safety regulation of the Nuclear Safety law is judged to be excessive regulation. Therefore, it is considered necessary to revise the regulations for radiation generators in the current Nuclear Safety law or to revise the radiation safety management system for university students.

• The Korean Journal of Nuclear Medicine Technology
• /
• v.17 no.1
• /
• pp.3-6
• /
• 2013

Purpose: The treatment of thyroid cancer patients was continuously increased. According to the increment of thyroid cancer patients, the establishment of iodine therapy site was also increased in each hospital. This treatment involves the administration of radioactive iodine, which will be given in the form of a capsule. Therefore, protections and managements for radioactive source pollution and radiation exposure should be necessary for radiation safety. Among the many problems, the problem of disposing the radioactive wastes was occurred. In this study, The date for self-disposal for radioactive wastes, which were contaminated in clothes, bedclothes and trash, were calculated. Materials and Methods: The number of iodine therapy ward was 15 in Korea Institute of Radiological Medical and Sciences. Recently, 8 therapy wards were operated for iodine therapy patients and others were on standby for emergency treatment ward of any radiation accidents. Radioactive wastes, which were occurred in therapy ward, were clothes, bedclothes, bath cover for patients washing water and food and drink which was leftover by patients. Each sample was hold into the marinelli beaker (clothes, bedclothes, bath covers) and 90 ml beaker (food, drink, and washing water). The activities of collected samples were measured by HpGe MCA device (Multi Channel Analysis, CANBERRA, USA) Results: The storage period for the each kind of radioactive wastes was calculated by equation of storage periods based on the measurement outcomes. The average storage period was 60 days for the case of clothes, and the maximum storage period was 93 days for patient bottoms. The average storage period and the maximum storage period for the trash were 69 days and 97 days, respectively. The leftover foods and drinks had short storage period (the average storage period was 25 days and maximum storage period was 39 days), compared with other wastes. Conclusion: The proper storage period for disposing the radioactive waste (clothes, bedclothes and bath cover) was 100 days by the regulation on self-disposal of radioactive waste. In addition, the storage period for disposing the liquid radioactive waste was 120 days. The current regulation for radioactive waste self-disposing was not suitable for the circumstances of each radioactive therapy facility. Therefore, it was necessary to reduce the leftover food and drinks by adequate table setting for patients, and improve the process and regulation for disposing the short-half life radioactive wastes.

• The Korean Journal of Nuclear Medicine Technology
• /
• v.15 no.2
• /
• pp.104-110
• /
• 2011

Purpose: The object evaluation method about medical institutes of these days increases credibility of consumers about medical services by conducting a certification system about medical institutes. In addition, as nuclear medicine test rooms and diagnosis test medicine room adopt many kinds of international certification systems, the matters regarding safety management of test rooms are being regarded as important. Since the blood test rooms of nuclear medicine are exposed to many harmful factors such as infection from clinical specimen and radioactive isotope reagent, there is a need to pay lots of attention to the safety management of staff and patients. Therefore, this study discusses safety management activities of staff and patients, which are conducted in the blood test rooms of the nuclear medicine department in Asan Medical Center. Materials and Methods: In the blood test rooms of the nuclear medicine department in Asan Medical Center, the matters regarding comprehensive safety management by the person in charge of safety management are offered and all staff members of the test rooms apply them into work. Safety management education is regularly conducted according to established regulations, and infection is prevented through implementation of wearing personal protectors and hand sanitation during test work. In addition, technical safety guides and accident guides for interruption of electric power are provided against emergencies. Through infection management guides, infection prevention and preparation methods for infection are learned and radioactive isotope management, safety management about reagent use and safety guides about harmful chemical substances are being applied to work. Results: The blood test rooms of the nuclear medicine department apply safety management regulations to work. Under the situation where hand sanitation should be conducted, hands are washed to prevent infection between staff and patients, and for preventing infection from clinical specimen, personal protectors are worn. The reagent, which is classified as harmful substance, is separately stored to be easily recognized, radioactive wastes and general medical wastes are also safely managed. Through these lots of safety management activities, safety management awareness of staff members is enhanced, and patients are protected from many dangers. Conclusion: Staff members of the blood test rooms of the nuclear medicine department should be fully aware of safety management regulations and apply them to work. When better safety management suggestions are made, they need to be examined and applied for increasing quality of safety management for staff and patients.

• Nuclear Engineering and Technology
• /
• v.54 no.12
• /
• pp.4499-4513
• /
• 2022

Spent fuel behavior of dry storage was simulated in a continuous state from steady-state operation by modifying FRAPCON-4.0 to incorporate spent fuel-specific fuel behavior models. Spent fuel behavior of a typical PWR was compared with that of NuScale Power Module (NPM^TM). Current PWR discharge burnup (60 MWd/kgU) gives a sufficient margin to the hoop stress limit of 90 MPa. Most hydrogen precipitation occurs in the first 50 years of dry storage, thereby no extra phenomenological safety factor is identified for extended dry storage up to 100 years. Regulation for spent fuel management can be significantly alleviated for LWR-based SMRs. Hydride embrittlement safety criterion is irrelevant to NuScale spent fuels; they have sufficiently lower plenum pressure and hydrogen contents compared to those of PWRs. Cladding creep out during dry storage reduces the subchannel area with burnup. The most deformed cladding outer diameter after 100 years of dry storage is found to be 9.64 mm for discharge burnup of 70 MWd/kgU. It may deteriorate heat transfer of dry storage by increasing flow resistance and decreasing the view factor of radiative heat transfer. Self-regulated by decreasing rod internal pressure with opening gap, cladding creep out closely reaches the saturated point after ~50 years of dry storage.

• Nuclear Engineering and Technology
• /
• v.49 no.8
• /
• pp.1772-1777
• /
• 2017

Korean nuclear power plants (NPPs) periodically evaluate the radioactive gaseous and liquid effluents released from power reactors to protect the public from radiation exposure. This paper provides a comprehensive overview of the release of radioactive effluents from Korean NPPs and the effects on the annual radiation doses to the public. The amounts of radioactive effluents released to the environment and the resulting radiation doses to members of the public living around NPPs were analyzed for the years 2011-2015 using the Korea Hydro & Nuclear Power Co., Ltd's annual summary reports of the assessment of radiological impact on the environment. The results show that tritium was the primary contributor to the activity in both gaseous and liquid effluents. The averages of effective doses to the public were approximately on the order of $10^{-3}mSv$ or $10^{-2}mSv$. Therefore, even though Korean NPPs discharged some radioactive materials into the environment, all effluents were within the regulatory safety limits and the resulting doses were much less than the dose limits.