Helium-cooled Fast Reactor Research Articles

Alternatives to the burnup strategy and fuel material of the EM2 modular reactor

The Energy Multiplier Module (EM2) is a modular high-temperature helium-cooled fast reactor developed by General Atomics (GA). It contemplates a convert-and-burn core, in which fertile fuel is converted to fissile fuel for on-site burning. These features allow the reactor to achieve a core lifetime of 32 years without refueling with different types of fuel materials, like thorium, natural uranium, depleted uranium (DU), and spent fuel. The objective of this study is to evaluate the neutronic performance of the EM2 reactor under four types of fuel material combinations (Low enriched uranium [LEU]/DU, LEU/Th, U-Pu/DU and U/Th) with two burnup strategies. The first strategy was taken from the GA model (configuration 1) and the second was proposed in this paper (configuration 2). The Serpent Monte Carlo code with ENDF/B-VII cross section library was used to perform the calculations. The following parameters were analyzed for each scenario: the effective neutron multiplication factor (keff), the conversion ratio, reactivity coefficients, and the effective delayed neutron fraction (βeff). The fissile inventory and fission rate contribution of the main isotopes were also discussed. The results obtained show that the scenarios with configuration 2 had a higher peak excess reactivity than those with configuration 1. For the scenarios with DU as a fertile fuel, 238U produced ∼20 % of the fission throughout the fuel cycle. In contrast, in the scenarios with thorium as fertile fuel, the 232Th fission rate was negligible. As for the reactivity coefficients, all scenarios presented negative Doppler coefficient and positive void reactivity; both values increased along burnup. On the other hand, the βeff decreased with burnup on and did not vary significantly between both configurations.

Study of Helium Cooled Fast Reactor Core Design Fuelled by Thorium Carbide-Uranium Carbide with Modified Candle Axial Direction Scheme

Human need of energy will increase time to time. Therefore, a safe, renewable, and efficient source of energy, which is Nuclear Energy, is needed. Nuclear Power Plant (NPP) is the most compatible solution to provide electricity to human race in the future. The problem that came within NPP is the danger of proliferation issues. The method that has been developed to overcome this problem is CANDLE [5] and has been modified by Prof. Zaki Su’ud (Modified CANDLE scheme). This research use Axial Modified CANDLE Scheme to Helium-Cooled Fast Reactor with Natural Uranium Carbide-Thorium Carbide as fuel and applied to various size of core as optimization. Neutronic aspect such as, burn up level, multiplication factor, and conversion ratio are utilized in this paper in order to analyse the behaviour of the reactor. Other than that, percentage of Uranium has been varied to reduce power peaking. The neutronic calculation has been done using SRAC and core design calculation by FI-ITB-CH1. This research concludes that power peaking reduction is able to achieve by combining Uranium Carbide and Thorium Carbide to the fuel. The optimum reactor design reached at 360 cm of core radius and 303 cm of core height.

Role of the Nonrotating Decay Heat Removal Blower Pressure Loss in ALLEGRO Reactor

Abstract ALLEGRO is a helium-cooled fast reactor, which is being developed by the Czech Republic, France, Hungary, Slovakia, and Poland. It will be the demonstrator of the GFR-2400-MWth Generation IV gas-cooled fast reactors. In ALLEGRO, a three-loop safety system is designed to remove decay heat during accident conditions. The nonrotating blower blades may represent a huge pressure loss in the decay heat removal loop (DHR), which hinders natural circulation. The lower the pressure loss coefficient of the DHR blower blades is, the better cooling is available during natural circulation. On the other hand, a large core bypass develops if a DHR valve is opened inadvertently during normal operation. In this case, the higher DHR blower pressure loss is better from core cooling point of view. Consequently, the low pressure loss of the DHR blower is advantageous for core cooling in station blackout (SBO) event but disadvantageous for inadvertent DHR valve opening event. Both the above-mentioned cases may lead to insufficient core cooling in accident conditions, which threatens the integrity of the reactor core. In this study, we present CATHARE thermohydraulic calculations to assess the sensitivity of the DHR blower pressure loss coefficient for the above-mentioned two cases.

Extension and application of the KIKO3DMG nodal code for fast reactor core analyses

The KIKO3DMG is a nodal reactor-physics calculation tool. This paper presents extensions, which were necessary to prepare the code for simulation of fast reactor transients. A new axial expansion module was developed for the KIKO3DMG, which introduces mixed nodes. The theoretical background and a new method of group constant mixing are described. Weighting the group constants with the volume ratios is appropriate for small, non-absorber nodes, if the inverses of the diffusion coefficients are weighted instead of the diffusion coefficients. Group constants were generated for the ALLEGRO helium-cooled fast reactor using the Serpent Monte Carlo code and applied in the KIKO3DMG. The effect of the group constant parameterization was investigated on the power distribution and the coefficients of reactivity. The axial and radial power distribution and the fuel, cladding and wrapper temperature coefficients of reactivity and the void coefficient of reactivity were calculated with good accuracy in the KIKO3DMG.

ALLEGRO Gas-cooled Fast Reactor (GFR) demonstrator thermal hydraulic benchmark

The European gas cooled fast reactor (GFR) demonstrator, ALLEGRO, is currently being developed by Czech Republic, France, Hungary, Slovakia and Poland. The primary aim of ALLEGRO is to demonstrate the helium-cooled fast reactor technology and to test the new ceramic refractory fuel of the industrial version of the gas-cooled fast reactor, which has an envisaged thermal power of 2400 MW. The development of ALLEGRO highly relies on Thermal Hydraulic (TH) codes which were primarily developed for two-phase flow modelling of pressurized and boiling water reactors, and which are currently being adapted for new generation reactors using gas or liquid metal coolant. In this paper we present CATHARE2, RELAP5-3D and MELCOR code to code benchmark carried out for the 75 MWth ALLEGRO reactor. During our work, first an ALLEGRO database for TH analyses was elaborated based on previous European projects. In the second step a detailed TH benchmark specification was prepared in order to help the TH model development as a third step. Fourthly, each participant calculated their own steady states and subsequently we identified several model distortions. In the final step we carried out transient calculations for 3-inch Loss Of Coolant Accident (LOCA) and total Station Blackout (SBO) initiating events. In general, the results show that there is a good agreement in the simulation results of the different codes, but further experimental code validations are proposed using helium-cooled experimental facilities, namely the S-ALLEGRO being built in Czech Republic and the STU (Slovak University of Technology) loop operated in Slovakia.

The Energy Multiplier Module (EM2): Status of Conceptual Design

The Energy Multiplier Module (EM2) is a helium-cooled fast reactor with a core outlet temperature of 850°C. It is designed as a modular, grid-capable power source with a net unit output of 265 MWe. The reactor employs a convert-and-burn core design that converts fertile isotopes to fissile and burns them in situ over a 30-year core life. The reactor is sited in a below-grade sealed containment. It uses passive safety methods for heat removal and reactivity control to protect the integrity of the fuel, reactor vessel, and containment. The plant also incorporates a below-grade, passively cooled spent fuel storage facility with capacity for 60 years of full-power operation. EM2 employs a direct closed-cycle gas turbine power conversion unit (PCU) with an organic Rankine bottoming cycle for 53% net power conversion efficiency assuming evaporative cooling. The high-power conversion efficiency and long-burn fuel cycle reduce the electricity cost by 35% when compared with the conventional light water reactor.The conceptual design has been conducted for the EM2 plant with focus on the reactor, fuel, and safety system designs. A detailed model of the passive direct reactor auxiliary cooling system was created to demonstrate functionality for selected design-basis accidents. The bench-scale fuel development campaign demonstrated high-quality uranium carbide pellet fabrication as well as β-SiC composite cladding and SiC-joining technologies. Irradiation tests of reactor materials are also being conducted. The PCU variable-speed generator mechanical design was validated with operational testing of its novel rotor at speeds >13 000 rpm. The design of the turbo-compressor with active magnetic bearings continues. A large cost database and financial model have been constructed for use as a key driver for the design to be economically competitive with competing generating technologies after 2030.

The Energy Multiplier Module: Advancing the Nuclear Fuel Cycle through Technology Innovations

To achieve long-term energy security in an environmentally acceptable manner, fission technology needs to make further advances in the areas of lower financial risk, better resource utilization, and reduced volumes of high-level waste. Without such progress, these concerns may be limiting factors in the exploitation of this vital resource. “Convert-and-burn” fast reactors offer the potential for advances in each of these areas without the specter of increased proliferation risk that accompanies breeder reactor concepts. An example is Energy Multiplier Module (EM2), a compact, helium-cooled fast reactor that augments its fissile fuel load with either depleted uranium or used nuclear fuel (UNF). The convert-and-burn in situ operating mode results in a core predicted to last 30 years without the need to add or shuffle fuel. EM2 can endure a station blackout, even one combined with a loss-of-coolant accident, using only passive safety systems to prevent radioactivity release or loss of plant. The end-of-cycle fuel and/or light water reactor UNF can be refabricated in a manner that does not separate out heavy metal, permitting reuse in subsequent generations at reduced proliferation risk. Proliferation resistance is further enhanced by eliminating the need for enrichment beyond that needed for the first-generation fuel load. Waste problems are mitigated by several factors: higher burnup, fuel use in multiple generations, and conversion of existing waste to energy. Economically attractive power costs are anticipated through a combination of high efficiency, simplicity of the direct-cycle gas turbine, and relatively small subsystems that can be shop fabricated and shipped by road to the site. Reactor materials have been carefully chosen to achieve a safe, economically affordable, and proliferation-resistant energy source.

Risk-Informed Design Changes in a Passive Decay Heat Removal System

The failure probability of a passive decay heat removal (DHR) system after a loss-of-coolant accident (LOCA) is evaluated as part of a risk-informed design process for a helium-cooled fast reactor. The system was modeled using RELAP5-3D. The epistemic uncertainties in input parameters as well as the epistemic model uncertainties in the code were assessed and propagated through the model using Latin hypercube sampling. The changes in the design that we investigated reduced the overall failure probability of the system by reducing the impact of the major contributor to the failure probability. Sensitivity analyses led to two unexpected results. First, the key factors affecting the system failure probability are the location of the thermal insulation (inside or outside the hot leg) and the uncertainty in the insulation thermal conductivity. Second, the heat transfer coefficient in the core is not as important as one might expect. Our results show that the heat transfer coefficient in the containment structures is more important. Different methods for sensitivity analysis were applied and gave consistent results. The calculated conditional (given a LOCA) failure probability of the passive DHR system was deemed to be unacceptable and led the Massachusetts Institute of Technology design team to adopt an active DHR system as the main mode of DHR for the gas-cooled fast reactor

Incorporating reliability analysis into the design of passive cooling systems with an application to a gas-cooled reactor

A time-dependent reliability evaluation of a two-loop passive decay heat removal (DHR) system was performed as part of the iterative design process for a helium-cooled fast reactor. The system was modeled using RELAP5-3D. The uncertainties in input parameters were assessed and were propagated through the model using Latin hypercube sampling. An important finding was the discovery that the smaller pressure loss through the DHR heat exchanger than through the core would make the flow to bypass the core through one DHR loop, if two loops operated in parallel. This finding is a warning against modeling only one lumped DHR loop and assuming that n of them will remove n times the decay power. Sensitivity analyses revealed that there are values of some input parameters for which failures are very unlikely. The calculated conditional (i.e., given the LOCA) failure probability was deemed to be too high leading to the identification of several design changes to improve system reliability. This study is an example of the kinds of insights that can be obtained by including a reliability assessment in the design process. It is different from the usual use of PSA in design, which compares different system configurations, because it focuses on the thermal–hydraulic performance of a safety function.

A Subcritical, Helium-Cooled Fast Reactor for the Transmutation of Spent Nuclear Fuel

A design concept and supporting analysis are presented for a He-cooled fast reactor for the transmutation of spent nuclear fuel. Coated transuranic (TRU) fuel particles in a SiC matrix are used. The reactor operates subcritical (k ≤ 0.95), with a tokamak D-T fusion neutron source, to achieve >90% TRU burnup in repeated five-batch fuel cycles, fissions 1.1 tonnes/full-power year, and produces 700 MW(electric) net electrical power. The reactor design is based on nuclear, fuels, materials, and separations technologies being developed in the Generation-IV, Next Generation Nuclear Plant, and Advanced Fuel Cycle Initiative programs and similar international programs, and the fusion neutron source is based on the physics and technology supporting the ITER design.

High temperature helium-cooled fast reactor (HTHFR)

Abstract Scoping calculations have been performed for a very high temperature (1000 °C) helium-cooled fast reactor involving two distinct options: (1) using graphite foam into which UC (12% enrichment) is embedded into a matrix comprising UC and graphite foam molded into hexagonal building blocks, shown in Fig. 1, and encapsulated with a SiC shell covering all surfaces, and (2) using UC only (also 12% enrichment) molded into the same shape and size as the foam–UC matrix in option 1. Both options use the same basic hexagonal fuel matrix blocks to form the core and reflector. The reflector contains natural uranium only. Both options use 50 μm SiC as a containment shell for fission product retention within each hexagonal block. The calculations show that the option using foam (option 1) would produce a reactor that can operate continuously for at least 25 years without ever adding or removing any fuel from the reactor. The calculations show further that the UC only option (option 2) can operate continually for 50 years without ever adding or removing fuel from the reactor. Doppler and loss of coolant reactivity coefficients were calculated. The Doppler coefficient is negative and much larger than the loss of coolant coefficient, which was very small and positive. Additional progress on and development of the two concepts are continuing.