Requirements on fuel management for a safe and optimum operation of the German research reactor FRJ-2

In case of a failure of a coarse control arm (CCAs) at FRJ-2, reactivity is added to the reactor. The amount of this reactivity depends on the efficiency of the individual CCAs which has been measured as 180% of the average reactivity of the six arms for the central arm. For this design basis accident, it is required that only four out of five residual arms are capable of shutting down the reactor. This minimum shutdown reactivity is provided by an optimum fuel management including an experimental reactivity determination. Calculation of fuel burnup and material densities is performed by the depletion code SUSAN, which has been verified by separate calculations using ORIGEN. The difference in the reactivity values (between calculation and measurement) is mainly a consequence of the limitation of the inverse kinetic method, which is incapable of covering the effects of the flux deformation and interaction of the CCAs and core in the process of reactor scram.     

The research reactor radiation emergency countermeasure system in Korea

A disaster prevention system was established for a radiation emergency from an operation of a research reactor with a thermal power of 30 MW_th in Korea. A national radiation disaster countermeasure organization was set up to cope with the radiation emergency classified into three cases whose effective doses were more than 1 mSv/h inside the nuclear facility, inside the site boundary and outside the site boundary. Its role consists of the proclamation and consequent withdrawal of a disaster, a general assessment, an emergency medical service, a field control, radiation protection, resident protection implement, an accident analysis, a security plan, a radiation environmental investigation plan and probe, a radiation environmental effect assessment, and others. The emergency planning zone (EPZ) was settled to be within a radius of 800 m, the average distance between the site boundary and the center of a research reactor in operation, as a quick and effective early countermeasure from the result of the radiation environmental effect assessment. The environmental probing zone was chosen to extend to a radius of 2 km from a research reactor according to the moving path of the radioactive cloud so that a densely populated area could be considered and would be extended to 10 km according to the radiation level of the research reactor and atmospheric diffusion. Practically, the environmental probing is implemented at 22 points inside the site and eight points outside the site considering the geography, population and the wind direction. The gamma radiation dose and atmospheric radioactivity are analyzed during an effluence, and the radioactivity of a ground surface deposit and an environmental sample are analyzed after an effluence. The environmental laboratory covers the analysis of the gamma radioisotopes, tritium, strontium, uranium, gross alpha and beta. It is estimated that the habitability can be recovered when the radiation dose rate is less than 1 mSv/h inside the site and around the environmental laboratory with the no sign of an effluence of the radioactive material. As a conclusion, it is thought that this emergency countermeasure system is effective in a real radiation emergency situation.     

An innovative research reactor design

A new and innovative core design for a research reactor is presented. It is shown that while using the standard, low enriched uranium as fuel, the maximum thermal flux per MW of power for the core design suggested and analyzed here is greater than those found in existing state of the art facilities without detrimentally affecting the other design specs. A design optimization is also carried out to achieve the following characteristics of a pool type research reactor of 10 MW power: high thermal neutron fluxes; sufficient space to locate facilities in the reflector; and an acceptable life cycle. In addition, the design is limited to standard fuel material of low enriched uranium. More specifically, the goal is to maximize the maximum thermal flux to power ratio in a moderate power reactor design maintaining, or even enhancing, other design aspects that are desired in a modern state of the art multi-purpose facility. The multi-purpose reactor design should allow most of the applications generally carried out in existing multi-purpose research reactors. Starting from the design of the German research reactor, FRM-II, which delivers high thermal neutron fluxes, an azimuthally asymmetric cylindrical core design with an inner and outer reflector, is developed. More specifically, one half of the annular core (0 < θ < π) is thicker than the other half. Two variations of the design are analyzed using MCNP, ORIGEN2 and MONTEBURNS codes. Both lead to a high thermal flux zone, a moderate thermal flux zone, and a low thermal flux zone in the outer reflector. Moreover, it is shown that the inner reflector is suitable for fast flux irradiation positions. The first design leads to a life cycle of 41 days and high, moderate and low (non-perturbed) thermal neutron fluxes of 4.2 × 10¹⁴ n cm⁻² s⁻¹, 3.0 × 10¹⁴ n cm⁻² s⁻¹, and 2.0 × 10¹⁴ n cm⁻² s⁻¹, respectively. Heat deposition in the cladding, coolant and fuel material is also calculated to determine coolant flow rate, coolant outlet temperature and maximum fuel temperature under steady-state operating conditions. Finally, a more compact version of the asymmetric core is developed where a maximum (non-perturbed) thermal flux of 5.0 × 10¹⁴ n cm⁻² s⁻¹ is achieved. The core life of this more compact version is estimated to be about 23 days.     

Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity (ρ_ex), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 × 10³Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.     

Design of the VVR-S research reactor

A water-cooled, water-moderated reactor for facilitating scientific research endeavors on applications of nuclear energy in peaceful pursuits has been built in the Soviet Union.Such reactors are currently completed and in operation in the Soviet Union and in other Socialist countries. Six such reactors were put into operation during 1957–1959; five reactors (four of which are built to handle power surges) are in the stage of preparation, assembly, and start-up tests.This article describes the design of the VVR-S reactor and its experimental facilities. The physical characteristics of the reactor have been described in an earlier paper [1].     

研究堆堆芯换料优化研究

利用特征统计算法(CSA)与先进格林函数节块法相结合,对研究堆平衡循环换料进行优化研究,以重水区热中子注量率最大为主要目标,得到了优于手工优化方案的平衡循环倒换料方案.     

Gas cooled fast reactor research in Europe

Research on the gas-cooled fast reactor system is directed towards fulfilling the ambitious long term goals of Generation IV (Gen IV), i.e., to develop a safe, sustainable, reliable, proliferation-resistant and economic nuclear energy system. In common with other fast reactors, gas-cooled fast reactors (GFRs) have exceptional potential as sustainable energy sources, for both the utilisation of fissile material and minimisation of nuclear waste through transmutation of minor actinides. The primary goal of GFR research is to develop the system primarily to be a reliable and economic electricity generator, with good safety and sustainability characteristics. However, for the longer term, GFR retains the potential for hydrogen production and other process heat applications facilitated through a high core outlet temperature which, in this case, is not limited by the characteristics of the coolant. In this respect, GFR can inherit the non-electricity applications of the thermal HTRs in a sustainable manner in a future in which natural uranium becomes scarce.GFR research within Europe is performed directly by those states who have signed the “System Arrangement” document within the Generation IV International Forum (the GIF), specifically France and Switzerland and Euratom. Importantly, Euratom provides a route by which researchers in other European states, and other non-European affiliates, can contribute to the work of the GIF, even when these states are not signatories to the GFR System Arrangement in their own right. This paper is written from the perspective of Euratom's involvement in research on the GFR system, starting with the 5th Framework Programme (FP5) GCFR project in 2000, through the FP6 project between 2005 and 2009 and looking ahead to the proposed activities within the current 7th Framework Programme (FP7). The evolution of the GFR concept from the 1960s onwards is discussed briefly, followed by the current perceived role, objectives and progress with the Generation IV GFR system.     

Second trip system for NRU research reactor

For the past four decades, the NRU research reactor has played an important role at the Chalk River Laboratories, Atomic Energy of Canada Limited, serving as one of its major research and isotope production facilities. To ensure that it continues as an effective facility, compliant with the current safety standards, a comprehensive upgrade program is underway. Adding a second trip system (STS) is part of this upgrade program, aiming at improving the effectiveness and reliability of the overall shutdown function. This document describes the main features and basic principles of the STS.The STS is an independent, seismically qualified trip system, that guarantees reactor shutdown even if the existing trip system fails. It is designed based on 2 out of 3 general coincidence logic, with minimal interferences and changes to the existing system. In addition to the manual trip in the main control room, a remote manual trip is provided in the new Qualified Emergency Response Centre, which is also seismically qualified and always accessible. Thus, for any reason, if the main control room becomes uninhabitable, the reactor still can be manually shut down from this centre.