MWth Gas-cooled Fast Reactor Research Articles

Neutronic comparative analysis between DOE and CEA model designs for fourth generation 2400 MW-th gas-cooled fast reactor

Neutronic comparative analysis between DOE and CEA model designs for fourth generation 2400 MW-th gas-cooled fast reactor

Comparison of uranium plutonium nitride (U-Pu-N) and thorium nitride (Th-N) fuel for 500 MWth gas-cooled fast reactor (GFR) long life without refueling

Summary Comparison of uranium plutonium nitride and thorium nitride fuel for 500 MWth gas-cooled fast reactor has been done. Gas-cooled fast reactor is one type of generation IV reactor that can be operated in high temperature. Due to the high temperature, it can be used in hydrogen production. In this study, we compare the neutronic analysis of two fuel types, ie, uranium nitride fuel (U,Pu)N and thorium nitride fuel (Th,U233)N. The neutronic calculation uses SRAC2006 code system, and the data libraries use JENDL4.0. First, the fuel pin calculation (PIJ calculation) has been done to take the macro data that are used in CITATION calculation. Both uranium and thorium fuel use heterogeneous configuration with 3 variation fuel in the core. F1 is located in the central core, F2 middle core, and F3 outer core. The variation of fuel fraction is 40% until 65%, cladding 10%, and coolant 25% until 40%. For (U,Pu)N fuel, the diameter of the active core is 220 cm, and the height of the active core is 110 cm. And for (Th,U233)N, the diameter of the active core is 250 cm and the height of the active core is 150 cm. The reflector radial-axial width is 50 cm. For uranium plutonium nitride fuel, the type of the fuel in one core is varied; ie, F1 is 8%, F2 is 10%, and F3 is 12%. For thorium nitride fuel, the type of the fuel in one core also varied; ie, F1 is 7.8%, F2 is 8%, and F3 is 8.8%. The optimum value of thorium nitride is when fuel fraction of region F1 = 60%, F2 = 57.5%, F3 = 60%, the burn up time up to 20 years without refueling, max k-eff value is 1.0109, and max excess reactivity value is 1.08%. Neutronic analysis shows that both uranium and thorium fuel have excess reactivity value less than 2% but thorium fuel has excess reactivity less than uranium nitride fuel. Uranium fuel has better breeding capability than thorium fuel. Therefore, it is better to use uranium fuel for fast reactor like GFR, which has high breeding capability.

WITHDRAWN: Comparison of uranium plutonium nitride (U[sbnd]Pu[sbnd]N) and thorium nitride (Th[sbnd]N) fuel for 500 MWth Gas Cooled Fast Reactor (GFR) longlife without refueling

WITHDRAWN: Comparison of uranium plutonium nitride (U[sbnd]Pu[sbnd]N) and thorium nitride (Th[sbnd]N) fuel for 500 MWth Gas Cooled Fast Reactor (GFR) longlife without refueling

Comparative Study on Various Geometrical Core Design of 300 MWth Gas Cooled Fast Reactor with UN-PuN Fuel Longlife without Refuelling

Nuclear power has progressive improvement in the operating performance of exiting reactors and ensuring economic competitiveness of nuclear electricity around the world. The GFR use gas coolant and fast neutron spectrum. This research use helium coolant which has low neutron moderation, chemical inert and single phase. Comparative study on various geometrical core design for modular GFR with UN-PuN fuel long life without refuelling has been done. The calculation use SRAC2006 code both PIJ calculation and CITATION calculation. The data libraries use JENDL 4.0. The variation of fuel fraction is 40% until 65%. In this research, we varied the geometry of core reactor to find the optimum geometry design. The variation of the geometry design is balance cylinder; it means that the diameter active core (D) same with height active core (H). Second, pancake cylinder (D>H) and third, tall cylinder (D<H). The reflector radial-axial width is 50 cm. The power is 300 MWth. First calculation, we calculate survey parameter for UN-PuN fuel with fissile contain from Plutonium waste LWR for each geometry. The minimum power density is around 72 Watt/cc, and maximum power density 114 Watt/cc. After we calculate with various geometry core, when we use the balance geometry, the k-eff value flattest and more stable than the others.

Fuel Fraction Analysis of 500 MWth Gas Cooled Fast Reactor with Nitride (UN-PuN) Fuel without Refueling

Nuclear Power Plant (NPP) is one of candidates which can support electricity demand in the world. The Generation IV NPP has fourth main objective, i.e. sustainability, economics competitiveness, safety and reliability, and proliferation and physical protection. One of Gen-IV reactor type is Gas Cooled Fast Reactor (GFR). In this study, the analysis of fuel fraction in small GFR with nitride fuel has been done. The calculation was performed by SRAC code, both Pij and CITATION calculation. SRAC2002 system is a code system applicable to analyze the neutronics of variety reactor type. And for the data library used JENDL-3.2. The step of SRAC calculation is fuel pin calculated by Pij calculation until the data homogenized, after it homogenized we calculate core reactor. The variation of fuel fraction is 40% up to 65%. The optimum design of 500MWth GFR without refueling with 10 years burn up time reach when radius F1:F2:F3 = 50cm:30cm:30cm and height F1:F2:F3 = 50cm:40cm:30cm, variation percentage Plutonium in F1:F2:F3 = 7%:10%:13%. The optimum fuel fraction is 41% with addition 2% Plutonium weapon grade mix in the fuel. The excess reactivity value in this case 1.848% and the k-eff value is 1.01883. The high burn up reached when the fuel fraction is low. In this study 41% fuel fraction produce faster fissile fuel, so it has highest burn-up level than the other fuel fraction.

Design Study of 200MWth Gas Cooled Fast Reactor with Nitride (UN-PuN) Fuel Long Life without Refueling

Design study of 200 MWth Gas Cooled Fast Reactor with UN-PuN fuel long life without refueling has been done. GFR is one type reactor in Generation IV reactor system. It uses helium coolant and fast neutron spectrum. Helium is chemical inert, single phase and low neutron moderation. In this study the calculations are performed by using SRAC code with PIJ calculation for the fuel pin cell calculation and CITATION calculation for core calculation. The data libraries use JENDL 3.2. The variation fuel fractions are 50% until 60%. The diameter active core is 150 cm and the height active core is 100 cm. The reflector radial-axial width is 50 cm. The variation of the powers are 100 MWth up to 500 MWth. The high power causes the high k-eff value. The optimum design is reached when the power is 200 MWth, variation percentage Plutonium for fuel F1:F2:F3=9%:11%:13%. The comparation of fuel:cladding:coolant fraction = 55%:10%:35%. The cooling down time of Plutonium is nine months. The optimum k-eff value is 1.0142 with excess reactivity value 1.403%. The decay of Plutonium decrease k-eff value in the beginning of burn up.

Parametric analyses of partial sub-assembly blockages of the 50 MWth gas cooled experimental technology and demonstrator reactor (ETDR)

This paper presents a parametric study of thermal hydraulic and structural mechanic analyses of accidental blockages of the hottest sub-assembly of the 50 MWth gas-cooled fast reactor, ETDR. The blockage ratios were 60% and 100% of the sub-assembly flow cross-section located at the first row of grid spacers. Temperature profiles in the fuel and cladding were calculated as a function of time using computational fluid dynamics. The results were incorporated into finite elements analyses to evaluate thermal and mechanical stresses and strains in the cladding and fuel. 2D simulations with generalized plane strains were used in the structural analyses applying the maximum power density in the pellet as calculated by CFD. The thermal analyses showed that a 60% partial blockage increases the maximum cladding temperature by 130 °C within a time period of 50 s, whereas a full blockage will lead to clad melting (1320 °C) in about 8 s after accident initiation. In the finite element analyses several, mainly conservative, assumptions have been made to incorporate the phenomena occurring in the pellet–cladding interaction. The results of the finite element analyses represent a first study of the pellet–cladding mechanical interface under given transients, exploring the development of a methodology to be used for future analyses. The model has been verified to predict realistic contact stresses in line with analytical solutions during partial or full blockages. However, more elaborate 2D and 3D contact models, with creep and irradiation creep material models for both fuel and clad, together with parametric studies on friction coefficients and number of fuel fragments are foreseen for future work on failure criteria.

Level 1 probabilistic safety assessment to support the design of the CEA 2400 MWth gas-cooled fast reactor

As part of the design of a 4th generation reactor, the integration of safety in the early phase of the concepts is expected. To date, probabilistic insights are increasingly employed in the safety demonstration in combination with the deterministic approach (e.g. to identify the sequences of complex failures, to justify the categorization of situations) and used, even at an early stage of design, to identify the reliability of systems and equipment to handle the safety objectives (expressed in terms of core damage frequency targets). Within this frame, the CEA has undertaken to assess the benefit of developing a probabilistic model to support the design of the 2400MWth gas-cooled fast reactor.In the building process of this level 1 probabilistic safety assessment, a first phase consisted in making a preliminary model that took only into account families of initiating events that were defined for the design of the decay heat removal dedicated loops, namely the loss of coolant accidents (representative of medium pressure situations) and loss of off-site power/station black-out transients (representative of high pressure situations). Owing to the results obtained with this preliminary L1PSA model, it emerged that an increased reliability of the DHR function in high pressure conditions (i.e. characterized by IEs not associated to the loss of integrity of the helium pressure boundary) is suitable to reduce the overall core damage frequency. The track was therefore chosen to require the use of normal loops as first line of provision of the DHR function, possibly including components or particular operating modes related to the secondary and tertiary circuits. In addition, this final L1PSA model is characterized by success criteria based on transient calculations performed with the CATHARE2 code and to a perimeter extended to all representative internal IEs at full operating power.This paper presents the building process and the main results related to these two successive L1PSA models. Finally, useful insights are translated into GFR design improvements that are leading to an overall CDF at full operating power that satisfies nowadays with the probabilistic target defined for 3rd generation reactors, being at least the objective for 4th generation reactors.

Reliability Assessment of 2400 MWth Gas-Cooled Fast Reactor Natural Circulation Decay Heat Removal in Pressurized Situations

As the 2400 MWth gas-cooled fast reactor concept makes use of passive safety features in combination with active safety systems, the question of natural circulation decay heat removal (NCDHR) reliability and performance assessment into the ongoing probabilistic safety assessment in support to the reactor design, named “probabilistic engineering assessment” (PEA), constitutes a challenge. Within the 5th Framework Program for Research and Development (FPRD) of the European Community, a methodology has been developed to evaluate the reliability of passive systems characterized by a moving fluid and whose operation is based on physical principles, such as the natural circulation. This reliability method for passive systems (RMPSs) is based on uncertainties propagation into thermal-hydraulic (T-H) calculations. The aim of this exercise is finally to determine the performance reliability of the DHR system operating in a “passive” mode, taking into account the uncertainties of parameters retained for thermal-hydraulical calculations performed with the CATHARE 2 code. According to the PEA preliminary results, exhibiting the weight of pressurized scenarios (i.e., with intact primary circuit boundary) for the core damage frequency (CDF), the RMPS exercise is first focusing on the NCDHR performance at these T-H conditions.