PbBi-cooled direct contact boiling water small reactor

A design concept of PbBi cooled direct contact boiling water small fast reactor (PBWFR) has been formulated with some design parameters identified. Water is injected into hot PbBi above the core, and direct contact boiling takes place in chimneys. Boiling bubbles rise due to buoyancy effects, which works as a lift pump for PbBi circulation. The generated steam passes through separators and dryers for the removal of PbBi droplets, and then flows into turbines for the generation of electricity. The system pressure of 7 MPa is as the same as that of the conventional boiling water reactors (BWRs). The outlet steam is superheated by 10°C to avoid the accumulation of condensate on a PbBi free surface in the reactor vessel. The control rods are inserted from above, which is different from the original concept. This insertion was chosen since the seal of steam at the top of the reactor vessel is technically much easier than the seal of PbBi at the bottom of the reactor vessel. The electric power of 150 MWe may be the maximum which is practically possible as a small reactor with economic competitiveness to conventional LWRs. A two-region core is designed. A decrease in reactivity was estimated to be 1.5%dk/kk′ for 15 years. A fuel assembly has 271 fuel rods with 12.0 mm in diameter and 15.9 mm in pitch in a hexagonal wrapper tube. The design limit of cladding temperature is specified to be 650°C for compatibility of cladding material with PbBi. As a result, the PbBi core outlet temperature becomes 460°C. The PbBi temperature rise in the core is 150°C. The conditions of the secondary coolant steam are as the same as those of conventional BWRs with thermal efficiency of 33%. The core is designed to have the breeding ratio of 1.1 and the refueling interval of 15 years as a reactor with a long-life core. Direct heat exchangers (DHX), reactor vessel air cooling systems (RVACS) and guard vessel are designed.     

Increasing the power of the high temperature reactor module

To alleviate the economic problems of the modular pebble bed high temperature reactor, its design was modified in such a way that the power output was increased from 200 to 350 MWth. The core geometry was changed from cylindrical to annular, and the pressure vessel diameter was increased to 6.7 m. Control rods are placed in both the outer reflector and the graphite central column. In a safety analysis, loss of heat sink, loss of coolant and water ingress accident were examined. Reactor shutdown and decay heat removal take place passively, and the maximum fuel temperature stays theoreticallybelow 1600 °C, implying full retention of the fission products in the fuel elements. The central column has a diminishing effect on the positive reactivity effect of water ingress. A cost analysis shows that the specific investment costs of a four-module plant would decrease by 26% and the electricity generating costs would reduce by 19%.     

Thermo-mechanical analysis of supercritical pressure light water-cooled fast reactor fuel rod by FEMAXI-6 code

Thermo-mechanical behaviors of supercritical pressure light water cooled fast reactor (SWFR) fuel rod and cladding have been investigated by FEMAXI-6 (Ver.1) code with high enriched MOX fuel at elevated operating condition of high coolant system pressure (25 MPa) and high temperature (500 °C in core average outlet temperature). Fuel rod failure modes and associated fuel rod design criteria that is expected to be limiting in SWFR operating condition have been investigated in this fuel rod design study. Fuel centerline temperature is evaluated to be 1853 °C and fission gas release fraction is about 45% including helium production. Cumulative damage fraction is evaluated by linear life fraction rule with time-to-rupture correlation of advanced austenitic stainless steel. In a viewpoint of mechanical strength of fuel cladding against creep rupture and cladding collapse at high operation temperature, currently available stainless steels or being developed has a potential for application to SWFR. Admissible design range in terms of initial gas plenum pressure and its volume ratio are suggested for fuel rod design The stress ranges suggested by this study could be used as a preliminary target value of cladding material development for SWFR application.     

Development and characterization of the control assembly system for the large 2400 MWth Generation IV gas-cooled fast reactor

The present paper is related to the design and neutronic characterization of the principal control assembly system for the reference large (2400 MWth) Generation IV gas-cooled fast reactor (GFR), which makes use of ceramic–ceramic (CERCER) plate-type fuel-elements with (U–Pu) carbide fuel contained within a SiC inert matrix. For the neutronic calculations, the deterministic code system ERANOS-2.0 has been used, in association with a full core model including a European fast reactor (EFR)-type pattern for the control assemblies as a starting point. More specifically, the core contains a total of 33 control (control system device: CSD) and safety (diverse safety device: DSD) assemblies implemented in three banks. In the design of the new control assembly system, particular attention was given to the heat generation within the assemblies, so that both neutronic and thermal–hydraulic constraints could be appropriately accounted for. The thermal–hydraulic calculations have been performed with the code COPERNIC, significant coolant mass flow rates being found necessary to maintain acceptable cladding temperatures of the absorber pins.     

Natural Uranium Heavy Water Reactors for Desalination a Comparison Bltween the Fuel Pin and Fuel Cwijster Concepts

Studies on the potential use of nuclear desalination in Egypt have been conducted. The choice of reactor type and system concept was influenced by emphasis on relaxed technological conditions so as to fit local circumstances . Natural uranium heavy water reactors were found promising. Finer details of type of fuel, cladding, coolant, and system arrangement were left for comparative studies. The question of fuel type, fuel pin versus fuel cluster, was considered. Simplified analytical and computational techniques were adopted and results verified with published criticality measurements. It was found that the pin design would have higher breeding potentials while the cluster would provide simpler core arrangement especially if a pressure vessel design is chosen. Results are given for a typical 40 MWth project study showing core features and system characteristics.     

Subchannel analysis of fuel assemblies of a lead–alloy cooled fast reactor

Lead–alloy cooled fast reactor is one of the six Gen-IV reactors. It has many attractive features such as excellent natural circulation performance, better shielding against gamma rays or energetic neutrons and potentially reduced capital costs. A natural circulation lead–alloy cooled fast reactor with 10 MWth is under design in China (hereafter called LFR-10MW). Fuel assemblies thermal hydraulic analysis is of vital importance for a successful design. A subchannel analysis code with flow distribution model was used to carry out the thermal hydraulic analysis. This work briefly gave the thermal-hydraulic design for the LFR-10MW and analyzed the thermal-hydraulic characteristics under steady-state condition using the subchannel analysis code. Whole core analysis was performed to locate the hottest fuel assembly using the code. The hottest fuel assembly was analyzed to obtain the cladding temperature, fuel temperature and coolant velocity. The maximum cladding temperature, the maximum fuel center temperature and the maximum coolant velocity are all below the design constraints. These results imply that the thermal-hydraulic design of LFR-10MW is feasible.     

Preliminary study of parameter uncertainty influence on Pressurized Water Reactor core design

Sub-channel analysis can improve the accuracy of reactor core thermal design. However, the important initial parameters contain various uncertainties during reactor operation. In this work, the Sub-channel Analysis Code of Supercritical reactor (SACOS) code, which is also applicable for Pressurized Water Reactor (PWR), was used to study the coolant flow characteristic and fuel rod heat transfer characteristic of 1/8 assembly which has the maximum linear power density in 300 MWe PWR core firstly. Then the Wilks' method and Response Surface Method (RSM) were utilized to determine the influence of sub-channel input parameters uncertainties on the highest temperature of reactor core fuel rod and Minimum Departure from Nucleate Boiling Ratio (MDNBR). The results show that in the most conservative conditions, the maximum temperature of the fuel rod and MDNBR were 2167.4 °C and 1.08, respectively. Considering the uncertainties of assembly inlet flow rate, inlet coolant temperature and system pressure, the 95% probability values (with 95% confidence) of fuel rod maximum and MDNBR calculated using response surface methodology were 2144.0 °C and 1.6, while they were 2137 °C and 1.74 calculated by Wilks' approach. Results show that the uncertainty analysis methods can provide larger reactor design criteria margin to improve the economy of reactor. Furthermore, the code was developed to have the capacity to perform the uncertainty study of sub-channel calculation.     

Static design of a liquid-salt-cooled pebble bed reactor (LSPBR)

A renewed interest has been raised for liquid-salt-cooled nuclear reactors. The excellent heat transfer properties of liquid-salt coolants provide several benefits, like lower fuel temperatures, higher average coolant temperature, increased core power density and better decay heat removal, and thus higher achievable core power. In order to benefit from the on-line refueling capability of a pebble bed reactor, the liquid salt pebble bed reactor (LSPBR) is proposed. This is a high temperature pebble bed reactor with a fuel design similar to existing HTRs, but using a liquid-salt as coolant. In this paper, the selection criteria for the liquid-salt coolant are described. Based on its neutronic properties, LiF–BeF₂ (flibe) was selected for the LSPBR. Two designs of the LSPBR were considered: a cylindrical core and an annular core with a graphite inner reflector. Coupled neutronic thermal-hydraulic calculations were performed to obtain the steady state power distribution and the corresponding fuel temperature distribution. Calculations were performed to investigate the decay heat removal capability in a protected loss-of-forced cooling accident. The maximum allowable power that can be produced with the LSPBR is hereby determined.     

The feasibility of cooling heavy-water reactors with supercritical fluids

It appears technically feasible to use supercritical carbon dioxide as a coolant for a CANDU-type reactor. A new supercritical loop is proposed in which the reactor is cooled by a single-phase fluid pumped in a high density liquid-like state. The supercritical fluid-cooled reactor has the advantage of gas-cooled reactors of avoiding dryout, and of liquid-cooled reactors of low coolant-circulation power. By eliminating dryout, the maximum operating temperature of the fuel sheath can be increased to 1021°F (550°C) for existing Canadian fuel bundles, with a coolant exit temperature of 855°F (458°C) producing steam comparable to that of conventional fossil-fuel plants. Since the reactor coolant exit temperature from the steam generator may be as high as 280°F (138°C) low-pressure steam may also be produced. A new dual-reheat cycle is proposed with an ideal overall plant efficiency of 33%, comparable to the Pickering generating station.     

Safety design of Pb–Bi-cooled direct contact boiling water fast reactor (PBWFR)

In Pb–Bi-cooled direct contact boiling water small fast reactor (PBWFR), steam is generated by direct contact of feedwater with primary Pb–Bi coolant above the core, and Pb–Bi coolant is circulated by steam lift pump in chimneys. Safety design has been developed to show safety features of PBWFR. Negative void reactivity is inserted even if whole of the core and upper plenum are voided hypothetically by steam intrusion from above. The control rod ejection due to coolant pressure is prevented using in-vessel type control rod driving mechanism. At coolant leak from reactor vessel and feedwater pipes, Pb–Bi coolant level in the reactor vessel required for decay heat removal is kept using closed guard vessel. Dual pipes for feedwater are employed to avoid leak of water. Although there is no concern of loss of flow accident due to primary pump trip, feedwater pump trip initiates loss of coolant flow (LOF). Injection of high pressure water slows down the flow coast down of feedwater at the LOF event. The unprotected loss of flow and heat sink (ATWS) has been evaluated, which shows that the fuel temperatures are kept lower than the safety limits.     

Irradiation behavior of HTGR coated particle fuel at abnormally high temperature

Irradiation behavior of high temperature gas-cooled reactor (HTGR) coated particles under temperature transient conditions was investigated in accordance with a design-base accident scenario for HTTR, a 30 MWth HTGR under construction at JAERI. One of the scenarios predicts that the fuel temperature of the block-type fuel elements rises to abnormally high temperature by blocking a coolant channel with some foreign substance. For simulating this scenario the fuel compacts incorporating the coated particles were irradiated at normal temperature in three capsules, followed by temperature transient up to a maximum of approximately 2000°C. The post-irradiation examinations, including surface inspection, metrology, ceramography and a measurement of coated particle failure were applied to the fuel compacts to investigate the thermal-transient effect on the fuel integrity. Integrity of the fuel compact was also assessed by an estimation of tangential stress introduced into the compact by the temperature transient.     

Meeting fuel temperature limits in an HTR-Module Reactor during depressurized core heat-up

The thermal behaviour of an HTR-Module Reactor is discussed for the design basis event of core heat-up after fast depressurization taking into account the most unfavourable initial state and uncertainties of input data. The reactor is designed to retain its fission products inside the fuel coatings even if all active components for decay heat removal and reactivity control should fail. To meet this goal maximum fuel temperatures during core heat-up should not exceed the technological limit of 1620°C, for which the integrity of the fuel coatings has been proven experimentally.Two-dimensional thermal-hydraulic calculations show that the maximum fuel temperature during core heat-up is expected to be 1472°C taking into account nominal full power operation as an initial state, a sudden depressurization in the beginning of the event, and nominal input data. The most unfavourable initial state is the steady state operation close to the scram set points, i.e. 105% power and increased cold and hot gas temperatures. Accounting for this leads to a maximum fuel temperature of 1522°C. Relevant uncertainties of input data are those of decay heat production, power distribution and core thermal conductivity and specific heat capacity. Their individual standard deviations can be combined to an integral uncertainty margin of ±86 K which covers two standard deviations. Hence the maximum fuel temperature taking into account unfavourable initial state and uncertainties is 1608°C.     

CAST3M/ARCTURUS: A coupled heat transfer CFD code for thermal–hydraulic analyzes of gas cooled reactors

The safety of gas cooled reactors (High Temperature Reactors (HTR), Very High Temperature Reactors (VHTR) or Gas Cooled Fast Reactors (GFR)) must be ensured by systems (active or passive) which maintain loads on component (fuel) and structures (vessel, containment) within acceptable limits under accidental conditions. To achieve this objective, thermal–hydraulics computer codes are necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Some key safety questions are related to the evaluation of decay heat removal and containment pressure and thermal loads. This requires accurate simulations of conduction, convection, thermal radiation transfers and energy storage. Coupling with neutronics is also an important modeling aspect for the determination of representative parameters such as neutronics coefficient (Doppler coefficient, Moderator coeffcient, …), critical position of control rods, reactivity insertion aspects, …. For GFR, the high power density of the core and its necessary reduced dimension cannot rely only on passive systems for decay heat removal. Therefore, forced convection using active safety systems (gas blowers, heat exchangers, …) are highly recommended. Nevertheless, in case of station black-out, the safety demonstration of the concept should be guaranteed by natural circulation heat removal. This could be performed by keeping a relatively high back-up pressure for pure helium convection and also by heavy gas injection. So, it is also necessary to model mixing of different gases, the on-set of natural convection and the pressure and thermal loads onto the proximate or guard containment. In this paper, we report on the developments of the CAST3M/ARCTURUS thermal–hydraulics (Lumped Parameter and CFD) code developed at CEA, including its coupling to the neutronics code CRONOS2 and the system code CATHARE. Elementary validation cases are detailed, as well as application of the code to benchmark problems such as the HTR-10 thermal–hydraulic exercise. Examples of containment thermal–hydraulics calculations for fast reactor design (GFR) are also detailed.     

Reactor core thermohydraulic transients with thermohydrodynamic coupling

A study of the reactor core thermohydraulics in an LMFBR has been performed for the strongly coupled thermo-hydrodynamic transients. A numerical method to solve the coupled energy-momentum equations among multichannels in a core is presented and the computer code ORIFS-TRANSIENT has been developed.The results of sample calculations for a flow coastdown transient to natural circulation following a reactor scram in a typical loop-type LMFBR are as follows: (1) the inter-subassembly coolant flow redistribution due to buoyancy forces is significant under the low flow condition, such as natural circulation; (2) the maximum coolant temperature was decreased by about 80°C (corresponding to about 22% in terms of hot channel factor) due to the flow redistribution; (3) due to thermohydrodynamic coupling between upper plenum and other regions, the maximum coolant temperature was decreased by about 9°C; (4) due to inter-subassembly heat redistribution, the maximum coolant temperature was increased by about 7°C.     

Core thermal–hydraulic design

The core thermal–hydraulic design for the HTTR is carried out to evaluate the maximum fuel temperature at normal operation and anticipated operation occurrences. To evaluate coolant flow distribution and maximum fuel temperature, we use the experimental results such as heat transfer coefficient, pressure loss coefficient obtained by mock-up test facilities. Furthermore, we evaluated hot spot factors of fuel temperatures conservatively.As the results of the core thermal–hydraulic design, an effective coolant flow through the core of 88% of the total flow, is achieved at minimum. The maximum fuel temperature appears during the high-temperature test operation, and reaches 1492 °C for the maximum through the burn-up cycle, which satisfies the design limit of 1495 °C at normal operation. It is also confirmed that the maximum fuel temperature at any anticipated operation occurrences does not exceed the fuel design limit of 1600 °C in the safety analysis.On the other hand, result of re-evaluation of analysis condition and hot spot factors based on operation data of the HTTR, the maximum fuel temperature for 160 effective full power operation days is estimated to be 1463 °C. It is confirmed that the core thermal–hydraulic design gives conservative results.     

Subchannel analysis of supercritical light water-cooled fast reactor assembly

Subchannel analyses have been carried out for supercritical water-cooled fast reactor fuel assembly. Peak cladding surface temperature difference arising from subchannel heterogeneities have been calculated by using the improved subchannel analysis code STARS and was evaluated to be about 18.5 °C. Several suggestions have been also made for reducing the PCST difference arising from channel heterogeneity. Influences of local power peaking on deflection of cladding surface temperature are explained with pin power distribution taken from core depletion calculation in this paper. Maximum cladding surface temperature at nominal condition is evaluated to be 645.3 °C over the cycle. Statistical thermal design uncertainty associated with PCST calculation is evaluated by Monte-Carlo sampling technique combined with subchannel analysis code. Maximum statistical design uncertainty of PCST is calculated to be 31 °C and is in a good agreement with that from RTDP method. Influence of downward flow in seed region on system sensitivity is investigated by improved Monte-Carlo thermal design procedure. Limiting thermal condition of MCST is 681 °C (650 °C of nominal + 31 °C) within 95/95 limit for SWFR.     

Thermal response of a modular high temperature reactor during passive cooldown under pressurized and depressurized conditions

The concept of inherent safety features of the modular HTR design with respect to passive decay heat removal through conduction, radiation and natural convection was first introduced in the German HTR-module (pebble fuel) design and subsequently extended to other modular HTR design in recent years, e.g. PBMR (pebble fuel), GT-MHR (prismatic fuel) and the new generation reactor V/HTR (prismatic fuel).This paper presents the numerical simulations of the V/HTR using the thermal-hydraulic code THERMIX which was initially developed for the analysis of HTRs with pebble fuels, verified by experiments, subsequently adopted for applications in the HTRs with prismatic fuels and checked against the results of CRP-3 benchmark problem analyzed by various countries with diverse codes.In this paper, the thermal response of the V/HTR (operating inlet/outlet temperatures 490/1000 °C) during post shutdown passive cooling under pressurized and depressurized primary system conditions has been investigated. Additional investigations have also been carried out to determine the influence of other inlet/outlet operating temperatures (e.g. 490/850, 350/850 or 350/1000 °C) on the maximum fuel and pressure vessel temperature during depressurized cooldown condition. In addition, some sensitivity analyses have also been performed to evaluate the effect of varying the parameters, i.e. decay heat, graphite conductivity, surface emissivity, etc., on the maximum fuel and pressure vessel temperature. The results show that the nominal peak fuel temperatures remain below 1600 °C for all these cases, which is the limiting temperature relating to radioactivity release from the fuel. The analyses presented in this paper demonstrate that the code THERMIX can be successfully applied for the thermal calculation of HTRs with prismatic fuel. The results also provide some fundamental information for the design optimization of V/HTR with respect to its maximum thermal power, operating temperatures, etc.     

Neutronic feasibility design of a small long-life HTR

Small high temperature gas-cooled reactors (HTRs) have the advantages of transportability, modular construction and flexible site selection. This paper presents the neutronic feasibility design of a 20 MWth U-Battery, which is a long-life block-type HTR. Key design parameters and possible reactor core configurations of the U-Battery were investigated by SCALE 5.1. The design parameters analyzed include fuel enrichment, the packing fraction of TRISO particles, the radii of fuel compacts and kernels, and the thicknesses of top and bottom reflectors. Possible reactor core configurations investigated include five cylindrical, two annular and four scatter reactor cores for the U-Battery. The neutronic design shows that the 20 MWth U-Battery with a 10-year lifetime is feasible using less than 20% enriched uranium, while the negative values of the temperature coefficients of reactivity partly ensure the inherent safety of the U-Battery. The higher the fuel enrichment and the packing fraction of TRISO particles are, the lower the reactivity swing during 10 years will be. There is an optimum radius of fuel kernels for each value of the fuel compact design parameter (i.e., radius) and a specific fuel lifetime. Moreover, the radius of fuel kernels has a small influence on the infinite multiplication factor of a typical fuel block in the range of 0.2–0.25 mm, when the radius of fuel compacts is 0.6225 cm and the lifetime of the fuel block is 10 years. The comparison of the cylindrical reactor cores with the non-cylindrical ones shows that neutron under-moderation is a basic neutronic characteristic of the reactor core of the U-Battery. Increasing neutron moderation by replacing fuel blocks with graphite blocks and dispersing the graphite blocks in the reactor core are two effective ways to increase the fuel burnup and lifetime of the U-Battery. Water or steam ingress may induce positive reactivity ranging from 0 to 0.16 Δk/k, which further demonstrates that the U-Battery is under-moderated.     

Passive heat removal by vessel cooling system of HTTR during no forced cooling accidents

The high temperature engineering test reactor (HTTR) being constructed by the Japan Atomic Energy Research Institute is a graphite-moderated, helium-cooled reactor with an outlet gas temperature of 950 °C.Two independent vessel cooling systems (VCSs) of the HTTR cool the reactor core indirectly during depressurized and pressurized accidents so that no forced direct cooling of the reactor core is necessary. Each VCS consists of a water cooling loop and cooling panels around the reactor pressure vessel (RPV). The cooling panels, kept below 90 °C, cool the RPV by radiation and natural convection and remove the decay heat from the reactor core during these accidents.This paper describes the design details and safety roles of the VCSs of the HTTR during depressurized and pressurized accidents. Safety analyses prove that the indirect core cooling by the VCSs and the inherent safety features of the reactor core prevent a temperature increase of the reactor fuel and fission product release from the reactor core during these conditions. Furthermore, it is confirmed that even if VCS failure is assumed during these accidents, the reactor core and RPV can remain in a safe state.     

Thermal hydraulic design of a liquid salt-cooled flexible conversion ratio fast reactor

A 2400 MWth liquid-salt cooled flexible conversion ratio reactor was designed, utilizing the ternary chloride salt NaCl-KCl-MgCl₂ (30-20-50%) as coolant. The reference design uses a wire-wrapped, hexagonal lattice core, and is able to achieve a core power density of 130 kW/l with a core pressure drop of 700 kPa and a maximum cladding temperature under 650 °C. Four kidney-shaped conventional tube-in-shell heat exchangers are used to connect the primary system to a 545 °C supercritical CO₂ power conversion system. The core, intermediate heat exchangers, and reactor coolant pumps fit in a vessel approximately 10 m in diameter and less than 20 m high. Lithium expansion modules (LEMs) were used to reconcile conflicting thermal hydraulic and reactor physics requirements in the liquid salt core. Use of LEMs allowed the design of a very favorable reactivity response which greatly benefits transient mitigation. A reactor vessel auxiliary cooling system (RVACS) and four redundant passive secondary auxiliary cooling systems (PSACSs) are used to provide passive heat removal, and are able to successfully mitigate both the unprotected station blackout transient as well as protected transients in which a scram occurs. Additionally, it was determined that the power conversion system can be used to mitigate both a loss of flow accident and an unprotected transient overpower.