Postulated Accident Conditions Research Articles

ORCA: A Tool for Radiological Consequences for Accidental Releases

Evaluating atmospheric dispersion and radiological doses in the vicinity of buildings is required for small modular reactors (SMRs) because of the reduced size of their exclusion area boundary. The current Canadian nuclear industry tool for these calculations implements the methodology defined in CSA Standard N288.2-M91, which was written to support large Canada Deuterium Uranium (CANDU) nuclear reactors as opposed to SMRs. The ORCA (On/offsite Radiological Consequences of Accidents) code has been developed to address this technical concern in addition to evaluating atmospheric dispersion and doses in the far field. The code calculates worker and public doses following an airborne release of radioactive material into the atmosphere under postulated accident conditions at a nuclear facility. The current paper presents the key assumptions and methods utilized in ORCA and discusses qualification of the software to the requirements of CSA Standard N286.7-16. The new model is applicable to SMRs and existing reactor designs and reduces conservatisms in the near field (i.e., <1 km from the source) relative to the methods in CSA N288.2-M91.

Preliminary study of the sub-medium São Francisco region overlooking the installation of future Brazilian nuclear power plants

Preliminary studies of site selection for the installation of nuclear plants in the northeast of Brazil have shown that the São Francisco river basin is a possible region for the implementation of these plants. For their operation, however, nuclear power plants require a reliable source of water for the cooling of heat rejected from the condenser that is sufficient for normal operation, shutdown, postulated accident conditions, and fire protection. The objective of this paper is to discuss safety issues and environmental impacts arising from the placement of nuclear power plants at the São Francisco river basin. The approach taken considered water availability throughout the year, the maintenance of thermal balance and climate changes. The present discussion indicates the necessity of a long-term management plan for the various uses of the São Francisco basin. An economic and social approach to the impacts of nuclear power plant operation and other agricultural and industrial activities in the region is also necessary. Finally, the study suggests the adoption of a variable-cycle cooling system, which will enable operation in all types of hydrological conditions expected in the São Francisco basin.

Experimental Study Of Helium Stratification In Multi-Compartment Containment Studies Facility

Abstract A significant amount of hydrogen may be released inside the containment of water cooled nuclear power reactor under postulated accident conditions. Its distribution in the multi-compartment containment geometry must be known to manage and mitigate the local hydrogen concentration in combustible pockets. An experimental study to characterize the behavior of a lighter gas (helium in place of hydrogen) in a multi-compartment Containment Studies Facility (CSF) has been pursued. Helium distribution experiments have been performed in CSF by varying important accident parameters like helium release rate, duration and injection area. The experimental studies performed in CSF depict helium stratification in the upper dome region. Stratification in terms of stratification/effective stratification factor determined for a range of experiments. The present experimental studies are important for understanding hydrogen distribution characteristics in multi-compartment containment geometry and benchmarking of CFD codes. Based on these studies some important prevailing practices for recombiner placement were endorsed.

Numerical Investigation into the Temperature Profile of a PHWR Channel Containing a Disassembled Fuel Bundle During a Postulated Accident Condition

A reactor core overheats due to decay heat generated in the fuel when an effective cooling medium is unavailable, such as in a loss-of-coolant accident combined with a loss of emergency core coolant. If the heat generated is not effectively dissipated, then at extreme temperatures, the structural strength of the bundle assembly may deteriorate, leading to slumping of fuel elements onto the inner wall of the pressure tube. It is essential to examine the temperature behavior of the channel containing fuel pins in a disassembled state in order to comprehend the impact of further thermally induced deformations in the channel during postulated accident conditions. Capturing the temperature of channel components at each circumferential position from experiments is extremely difficult; thus, a modeling tool is necessary to obtain a thorough circumferential temperature profile. This paper presents a numerical study that aims to study the temperature distributions in a 1-m-long pressurized heavy water reactor (PHWR) channel containing a disassembled fuel bundle. The channel geometry and the boundary conditions implemented were obtained from the experiment. A temperature profile for each channel element at every circumferential and axial location was obtained. A thorough comparison of the predicted and the reported experimental values was performed, and it was found that the predicted temperature behavior of the channel was consistent with the experimental data. Further simulations with different fuel element configurations and decay powers may be carried out; in addition, the results obtained may be used for coupled thermal-mechanical and thermal-mechanical-chemical simulations.

Development of a Multi-Compartment Containment thermal hydraulic code: PARIRODHAN

The development of a multi-compartment containment thermal–hydraulic code (named PARIRODHAN) is presented in this paper. It provides the capability to predict the containment thermal hydraulics along with steam and non-condensable transport inside a containment building that follows after the release of mass and energy from the heat transport system during postulated accident conditions. The code is generic and provides the flexibility to the user to decide the number of compartments, junctions and associated wall structures. The code uses a pressure implicit scheme to obtain a stable and computationally efficient solution. It has been verified by comparing the code calculation with numerical and experimental results. Some benchmarking studies largely affecting the containment thermal-hydraulics are also presented in the paper. The benchmarking results show the capability of the code in simulating containment behaviour under blowdown conditions.

Thermal-Hydraulic Safety Assessment of Full-Scale ESBWR Nuclear Reactor Design

Abstract The economic simplified boiling water reactor (ESBWR) is a boiling water nuclear reactor of generation III+. The U.S. Nuclear Regulatory Commission (NRC) approved the ESBWR design as the world's best light-water nuclear reactor in 2014. It has the lowest core damage frequency (industry standard indicator of safety) of any Generation III or III+ reactor. It can cool automatically for more than seven days without using electricity or human intervention. During the operation, the ESBWR is designed to produce electricity while emitting almost no greenhouse gases. The energy generated by an ESBWR will prevent the emission of approximately 7.5 million metric tons of CO2 per year compared to standard electricity production on the U.S. grid. The analysis present in this paper aimed to characterize the thermal-hydraulic simulations of full-scale ESBWR design. The analysis presented will help in recognizing the improvement needed in the reactor design and its passive safety systems. The analysis is performed for normal steady-state and postulated design basis accident scenarios (break in one of the reactor main-steam lines inside the contentment). The simulation results obtained by the code REALP/SCDAPSIM/MOD3.4 are compared with the TRACG04 and MELCOR2.1 code results to determine the code predictability and accuracy under accident conditions of the newly proposed design of the ESBWR nuclear reactor. A parametric study is also carried out with different in-containment break sizes to determine code results' sensitive nature due to natural circulation flow phenomena, which are highly unstable. The analysis shows that the code has produced reasonably accurate results under the tested system thermal-hydraulic conditions compared with the TRACG04 and MELCOR2.1 code simulation results. This agreement between the three codes simulation results demonstrated that the publicly available RELAP5 code model has adequate features to simulate the ESBWR thermal-hydraulics. It has been also demonstrated that for the postulated accident conditions, the design of passive safety systems is capable to capture the accident progression without any active power.

Experimental and Numerical Investigation into Temperature Distribution of a Simulated PHWR Coolant Channel Under Heatup Condition

The postulated dual-failure accident, i.e., loss of primary coolant flow along with impairment of the emergency coolant injection system, leads to peak fuel temperatures. It is well known that the temperature of the fuel assemblies is one of the significant factors that affect the outcome of an accident. Therefore, the present work aims to thoroughly investigate the thermal response of a single channel under postulated accident conditions. An experimental system was developed to capture the steady-state heat and temperature distribution in a representative 37-element fuel channel for a decay heat of 6.13 kW. Ohmic heating of the fuel rod simulators (FRSs) mimicked the generation of radioactive decay heat. Numerical simulation was also performed using the Fluent 19.1® code, and the discrete ordinates method was used to solve the radiative transfer equation. Based on the experimental results and the simulation results, it was found that the maximum Zircaloy-4 cladding temperature ≈850°C to 870°C was in the center ring. The temperature was found to vary around the circumference for each of the FRSs. Furthermore, the outer ring FRSs that had the lowest temperature developed the highest circumferential temperature gradient. In the pressure tube, the average circumferential temperature gradient obtained from the experiment and the simulation was 3.76°C/radian and 3.85°C/radian, respectively. Between the calandria tube and the moderator, the heat transfer coefficient was estimated to be around 822.3 W/m2‧K.

Design and operation of the transparent integral effect test facility, URI-LO for nuclear innovation platform

Conventional integral effect test facilities were constructed to enable the precise observation of thermal–hydraulic phenomena and reactor behaviors under postulated accident conditions to prove reactor safety. Although these facilities improved the understanding of thermal–hydraulic phenomena and reactor safety, applications of new technologies and their performance tests have been limited owing to the cost and large scale of the facilities. Various nuclear technologies converging 4th industrial revolution technologies such as artificial intelligence, drone, and 3D printing, are being developed to improve plant management strategies. Additionally, new conceptual passive safety systems are being developed to enhance reactor safety. A new integral effect test facility having a noticeable scaling ratio, i.e., the (UNIST reactor innovation loop (URI-LO), is designed and constructed to improve the technical quality of these technologies by performance and feasibility tests. In particular, the URI-LO, which is constructed using a transparent material, enables better visualization and provides physical insights on multidimensional phenomena inside the reactor system. The facility design based on three-level approach is qualitatively validated with preliminary analyses, and its functionality as a test facility is confirmed through a series of experiments. The design feature, design validation, functionality test, and future utilization of the URI-LO are introduced.

Nonlinear Acoustic Spectroscopy Method for Nondestructive Testing of Thermally Damaged Concrete

Most concrete structures are subjected to a range of temperature corresponding to normal environmental temperatures. However, there are important cases where concrete structures may be exposed to much higher temperatures (e.g., jet aircraft engine blasts, building fires, chemical and metallurgical industrial applications in which the concrete is in close proximity to furnaces, and some nuclear power-related postulated accident conditions). Exposure of concrete to high temperatures affects its mechanical properties. In this paper we examine the dependence of the fundamental frequency on temperature to which the concrete beams were heated. Fundamental frequencies were obtained by an innovative method used Pseudorandom Binary Sequence of Maximum Length as a perturbation signal. For the verification of the results the Ultrasonic Pulse Velocity in concrete were also measured and flexural bending strengths were determined. The results show method with Pseudorandom Binary Sequence of Maximum Length as a perturbation signal as a very promising for non-destructive testing of thermally damaged concrete.

Evaluation of the Enhanced LEU Fuel (ELF) Design for Conversion of the Advanced Test Reactor to a Low-Enrichment Fuel Cycle

A conceptual low-enrichment uranium (LEU) fuel design has been developed for the Advanced Test Reactor (ATR) at Idaho National Laboratory. The ATR is currently fueled with a high-enrichment fuel but is slated to be converted to LEU under programs led by the National Nuclear Security Administration of the U.S. Department of Energy. A conceptual LEU fuel design, the Enhanced LEU Fuel (ELF), has been developed assuming power peaking control through the use of variable fuel meat thicknesses and no use of burnable poison. In initial work, this design was shown to satisfy performance requirements for ATR operation. Following these design calculations, a safety analysis process was initiated to demonstrate that the ELF design would successfully meet safety limits for postulated accident conditions. Those calculations, performed using RELAP5 and ATR-SINDA, require physics analysis to provide spatial power distributions and kinetics parameters for various core operations configurations. This article describes the findings of the physics analysis and provides predictions for the behavior of a LEU-fueled version of ATR, and compares these to calculations of the performance of the current high-enrichment uranium fuel.

水平管での気液対向流制限に対する一次元計算の予測精度

Countercurrent flow limitation (CCFL) may occur in a hot leg (consisting of an inclined pipe, a vertical 50-deg elbow, and a horizontal pipe) and a pressurizer surge line (consisting of a vertical pipe, a vertical 90-deg elbow, and a slightly inclined pipe with elbows) under postulated accident conditions in a pressurized water reactor (PWR). In our previous study, we developed a one-dimensional (1D) computation method to predict CCFL characteristics in an inclined pipe with elbows, to generalize the prediction method for CCFL in piping systems. In this study, we did 1D computations for CCFL experiments in horizontal pipes (diameters of D = 0.03-0.65 m and ratios of the length to the diameter of L/D = 4.5-63) to evaluate prediction accuracy of the 1D computation method. The differences of the CCFL constant C (value of JG*1/2 at JL*1/2 = 0 where JG* and JL* are the Wallis parameters) between measured and computed values were within dC = ±0.055 for C = 0.4-0.7.

ROLE OF PASSIVE SAFETY FEATURES IN PREVENTION AND MITIGATION OF SEVERE PLANT CONDITIONS IN INDIAN ADVANCED HEAVY WATER REACTOR

Pressing demands of economic competitiveness, the need for large-scale deployment, minimizing the need of human intervention, and experience from the past events and incidents at operating reactors have guided the evolution and innovations in reactor technologies. Indian innovative reactor ‘AHWR’ is a pressure-tube type natural circulation based boiling water reactor that is designed to meet such requirements, which essentially reflect the needs of next generation reactors. The reactor employs various passive features to prevent and mitigate accidental conditions, like a slightly negative void reactivity coefficient, passive poison injection to scram the reactor in event of failure of the wired shutdown systems, a large elevated pool of water as a heat sink inside the containment, passive decay heat removal based on natural circulation and passive valves, passive ECC injection, etc. It is designed to meet the fundamental safety requirements of safe shutdown, safe decay heat removal and confinement of activity with no impact in public domain, and hence, no need for emergency planning under all conceivable scenarios. This paper examines the role of the various passive safety systems in prevention and mitigation of severe plant conditions that may arise in event of multiple failures. For the purpose of demonstration of the effectiveness of its passive features, postulated scenarios on the lines of three major severe accidents in the history of nuclear power reactors are considered, namely; the Three Mile Island (TMI), Chernobyl and Fukushima accidents. Severe plant conditions along the lines of these scenarios are postulated to the extent conceivable in the reactor under consideration and analyzed using best estimate system thermal-hydraulics code RELAP5/Mod3.2. It is found that the various passive systems incorporated enable the reactor to tolerate the postulated accident conditions without causing severe plant conditions and core degradation.

Anisotropic deformation of Zr–2.5Nb pressure tube material at high temperatures

Abstract Zr–2.5Nb alloy is used for the pressure tubes in CANDU® reactor fuel channels. In reactor, the pressure tube normally operates at 300 °C and experiences a primary coolant fluid internal pressure of approximately 10 MPa. Manufacturing and processing procedures generate an anisotropic state in the pressure tube which makes the tube stronger in the hoop (transverse) direction than in the axial (longitudinal) direction. This anisotropy condition is present for temperatures less than 500 °C. During postulated accident conditions where the material temperature could reach 1000 °C, it might be assumed that the high temperature and subsequent phase change would reduce the inherent anisotropy, and thus affect the deformation behaviour (ballooning) of the pressure tube. From constant-load, rapid-temperature-ramp, uniaxial deformation tests, the deformation rate in the longitudinal direction of the tube behaves differently than the deformation rate in the transverse direction of the tube. This anisotropic mechanical behaviour appears to persist at temperatures up to 1000 °C. This paper presents the results of high-temperature deformation tests using longitudinal and transverse specimens taken from as-received Zr–2.5Nb pressure tubes. It is shown that the anisotropic deformation behaviour observed at high temperatures is largely due to the stable crystallographic texture of the α-Zr phase constituent in the material that was previously observed by neutron diffraction measurements during heating at temperatures up to 1050 °C. The deformation behaviour is also influenced by the phase transformation occurring at high temperatures during heating. The effects of texture and phase transformation on the anisotropic deformation of as-received Zr–2.5Nb pressure tube material are discussed in the context of the tube ballooning behaviour. Because of the high temperatures in postulated accident scenarios, any irradiation damage will be annealed from the pressure tube material and thus the unirradiated material results presented in this paper are also applicable to irradiated pressure tubes.

Analytical studies of the heat removal capability of a passive auxiliary feedwater system (PAFS)

As passive safety features for nuclear power plants receive increasing attention, various studies have been conducted to develop safety systems for third-generation (GEN-III) nuclear power plants that are driven by passive systems, such as natural circulation, gravity, and resistance to high temperatures. Thus, South Korea has designed the Advanced Power Reactor Plus (APR+) with a two-loop PWR and 1500 MWe by adding passive safety features to the Advanced Power Reactor 1400 MWe (APR1400). The Passive Auxiliary Feedwater System (PAFS) is one of several passive safety systems being designed for the APR+, and extensive studies are being conducted to complete its design and to verify its feasibility. Because the PAFS removes decay heat from the reactor core under transient and accident conditions, it is necessary to evaluate the heat removal capability of the PAFS under hypothetical accident conditions. Therefore, in this paper, after introducing the characteristics of the PAFS and its design requirements, a performance analysis of the PAFS is performed for two accident cases: Loss of Condenser Vacuum (LOCV) and Feedwater Line Break (FLB). For the analysis, the Multi-dimensional Analysis of Reactor Safety (MARS) code is used, and a MARS model is developed by adding the PAFS model to the existing APR1400 model. The analysis results show that the PAFS has enough capacity to remove decay heat under the postulated accident conditions. In addition, the adequacy of modified control logic for main steam isolation valve (MSIV) is validated by comparing the traditional control logic.

Investigating the execution of EOPs in LOTDFP + ATWS + LOCA scenarios for an ABWR using MAAP code

The effectiveness of the execution of emergency operation procedures (EOPs) for an advanced boiling water reactor (ABWR) during postulated accident conditions using MAAP 4 code is discussed in this paper. The simulation scenarios included the loss of turbine driven feedwater pump (LOTDRFP), the anticipated transients without scram (ATWS), and the loss of coolant accident (LOCA). Based on the comparisons of responses on different parameters for cases with and without EOP actions, we concluded that the EOPs could effectively mitigate the consequences of the accidents. In addition, the emergency depressurization (ED) timing and the times spent between executing the EOP steps were also considered. The simulation results clearly reveal that both the earlier execution of ED and the decrease of times spent between each EOP step could delay the boron injection and leave the operator ample time to take some other remedy actions for reactor safe shutdown.

Theoretical modeling of condensation of steam outside different vertical geometries (tube, flat plates) in the presence of noncondensable gases like air and helium

Condensation of steam coming out from the coolant pipe during a loss of coolant accident (LOCA) plays a key role in removing heat from the primary containment of the advanced nuclear reactor (ANR). The presence of large mass fractions of air ( W air = 0.25–0.9) and a small mass fraction of helium ( W He = 0.017–0.083) reduces the overall heat transfer coefficient (HTC) substantially. The present work emphasizes on the issue that modeling the diffusion of water-vapor through the gas–liquid interface is the key to give good predictions in HTC. In this, condensation conductivity and effective diffusivity plays a key role. Therefore, modifications have been made in the derivation for calculation of condensation conductivity in the case of steam–air mixture and effective diffusivity (in the case of multicomponent mixture). The model validation has been done with the experimental data of Dehbi et al. [Dehbi, A.A., Golay, M.W., Kazimi, M.S., 1991. National Conference of Heat Transfer AIChE Symposium Series, pp. 19–28] and Anderson et al. [Anderson, M.H., Herranz, L.E., Corradini, M.L., 1998. Experimental analysis of heat transfer within the AP600 containment under postulated accident conditions. Nucl. Eng. Des. 185, 153–172] and other analytical models available in the literature [Herranz, L.E., Anderson, M.H., Corradini, M.H., 1998a. The effect of light gases in noncondensable mixtures on condensation heat transfer. Nucl. Eng. Des. 183, 133–150; Herranz, L.E., Anderson, M.H., Corradini, M.L., 1998b. A diffusion layer model for steam condensation within the AP600 containment. Nucl. Eng. Des. 185, 153–172; Peterson, P.F., Schrock, Y.E., Kageyama, T., 1993. Diffusion layer theory for turbulent vapor condensation with noncondensable gases. J. Heat Transf., 115; Dehbi, A.A., Golay, M.W., Kazimi, M.S., 1991. National Conference of Heat Transfer AIChE Symposium Series, pp. 19–28]. Since the validations of the results were found satisfactory, the datasets [Dehbi, A.A., Golay, M.W., Kazimi, M.S., 1991. National Conference of Heat Transfer AIChE Symposium Series, pp. 19–28; Anderson, M.H., Herranz, L.E., Corradini, M.L., 1998. Experimental analysis of heat transfer within the AP600 containment under postulated accident conditions. Nucl. Eng. Des. 185, 153–172] have been compared with a wide range of subcooling and the operating pressures. An extensive comparison has been reported and the results predicted by the present model were found to be satisfactory.

Validation of constraint‐based methodology in structural integrity of ferritic steels for nuclear reactor pressure vessels

ABSTRACTVOCALIST (validation of constraint‐based methodology in structural integrity) was a shared cost action project co‐financed by DG Research of the European Commission under the Fifth Framework of the European Atomic Energy Community (EURATOM). The motivation for VOCALIST was based on the understanding that the pattern of crack‐tip stresses and strains causing plastic flow and fracture in components is different to that in test specimens. This gives rise to the so‐called constraint effect. Crack‐tip constraint in components is generally lower than in test specimens. Effective toughness is correspondingly higher. The fracture toughness measured on test specimens is thus likely to underestimate that exhibited by cracks in components. The purpose of VOCALIST was to develop validated models of the constraint effect and associated best practice advice, with the objective of aiding improvements in defect assessment methodology for predicting safety margins and making component lifetime management decisions. The main focus in VOCALIST was an assessment of constraint effects on the cleavage fracture toughness of ferritic steels used in the fabrication of nuclear reactor pressure vessels, because of relevance to the development of improved safety assessments for plant under postulated accident conditions. This paper provides a detailed summary of the main results and conclusions from VOCALIST and points out their contribution to advances in constraint‐based methodology for structural integrity assessment. In particular, the output from VOCALIST has improved confidence in the use of KJ−Tstress and KJ−Q approaches to assessments of cleavage fracture where the effects of in‐plane constraint are dominant. Cleavage fracture models based on the Weibull stress, σW, have been shown to be reliable, although current best practice advice suggests that σW should be computed in terms of hydrostatic stress (as distinct from maximum principal stress) for problems involving out‐of‐plane loading. Correspondingly, the results suggest that the hydrostatic parameter, QH, is the appropriate one with which to characterize crack‐tip constraint in analysing such problems. The materials characterization test results generated as part of VOCALIST have provided added confidence in the use of sub‐size specimens to determine the Master Curve reference temperature, T0, for as‐received and degraded ferritic RPV materials. The usefulness of correlating the Master Curve reference temperature, T0, with the constraint parameter, Q, has been demonstrated; however, the trend curves derived require further development and validation before they can be used in fracture analyses. The output from VOCALIST has contributed in providing the validation of methodology necessary to underpin the diffusion of constraint‐based fracture mechanics arguments in RPV safety cases, with potential applications including WWER as well as Western‐style LWR reactor types.

06/01626 Optimal design of the structure and settings of nuclear HTR thermodynamic cycles: Gosset, J. et al. International Journal of Thermal Sciences, 2005, 44, (12), 1169–1179

India has chalked out a nuclear power program based on its domestic resource position of uranium and thorium. The first stage started with setting up the Pressurized Heavy Water Reactors (PHWR) based on natural uranium and pressure tube technology. In the second phase, the fissile material base will be multiplied in Fast Breeder Reactors using the plutonium obtained from the PHWRs. Considering the large thorium reserves in India, the future nuclear power program will be based on thorium–233U fuel cycle. However, there is a need for the timely development of thorium-based technologies for the entire fuel cycle. The Advanced Heavy Water Reactor (AHWR) has been designed to fulfill this need. The AHWR is a 300 MWe, vertical, pressure tube type, heavy water moderated, boiling light water cooled natural circulation reactor. The fuel consists of (Th–Pu)O2 and (Th–233U)O2 pins. The fuel cluster is designed to generate maximum energy out of 233U, which is bred in situ from thorium and has a slightly negative void coefficient of reactivity. For the AHWR, the well-proven pressure tube technology has been adopted and many passive safety features, consistent with the international trend, have been incorporated. A distinguishing feature which makes this reactor unique, from other conventional nuclear power reactors is the fact that it is designed to remove core heat by natural circulation, under normal operating conditions, eliminating the need of pumps. In addition to this passive feature, several innovative passive safety systems have been incorporated in the design, for decay heat removal under shut down condition and mitigation of postulated accident conditions. The design of the reactor has progressively undergone modifications and improvements based on the feedbacks from the analytical and the experimental R&D. This paper gives the details of the current design of the AHWR.

Design and development of the AHWR—the Indian thorium fuelled innovative nuclear reactor

India has chalked out a nuclear power program based on its domestic resource position of uranium and thorium. The first stage started with setting up the Pressurized Heavy Water Reactors (PHWR) based on natural uranium and pressure tube technology. In the second phase, the fissile material base will be multiplied in Fast Breeder Reactors using the plutonium obtained from the PHWRs. Considering the large thorium reserves in India, the future nuclear power program will be based on thorium– 233U fuel cycle. However, there is a need for the timely development of thorium-based technologies for the entire fuel cycle. The Advanced Heavy Water Reactor (AHWR) has been designed to fulfill this need. The AHWR is a 300 MW e, vertical, pressure tube type, heavy water moderated, boiling light water cooled natural circulation reactor. The fuel consists of (Th–Pu)O 2 and (Th– 233U)O 2 pins. The fuel cluster is designed to generate maximum energy out of 233U, which is bred in situ from thorium and has a slightly negative void coefficient of reactivity. For the AHWR, the well-proven pressure tube technology has been adopted and many passive safety features, consistent with the international trend, have been incorporated. A distinguishing feature which makes this reactor unique, from other conventional nuclear power reactors is the fact that it is designed to remove core heat by natural circulation, under normal operating conditions, eliminating the need of pumps. In addition to this passive feature, several innovative passive safety systems have been incorporated in the design, for decay heat removal under shut down condition and mitigation of postulated accident conditions. The design of the reactor has progressively undergone modifications and improvements based on the feedbacks from the analytical and the experimental R&D. This paper gives the details of the current design of the AHWR.

The Effect of Pitch, Burnup, and Absorbers on a TRIGA Spent-Fuel Pool Criticality Safety

It has been shown that supercriticality might occur for some postulated accident conditions at the TRIGA spent-fuel pool. However, the effect of burnup was not accounted for in previous studies. In this work, the combined effect of fuel burnup, pitch among fuel elements, and number of uniformly mixed absorber rods for a square arrangement on the spent-fuel pool keff is investigated.The Monte Carlo computer code MCNP4B with the ENDF-B/VI library and detailed three dimensional geometry was used. The WIMS-D code was used to model the isotopic composition of the standard TRIGA and FLIP fuel for 5, 10, 20 and 30% burnup level and 2- and 4-yr cooling time.The results show that out of the three studied effects, pitch from contact (3.75 cm) up to rack design pitch (8 cm), number of absorbers from zero to eight, and burnup up to 30%, the pitch has the greatest influence on the multiplication factor keff. In the interval in which the pitch was changed, keff decreased for up to ~0.4 for standard and ~0.3 for FLIP fuel. The number of absorber rods affects the multiplication factor much less. This effect is bigger for more compact arrangements, e.g., for contact of standard fuel elements with eight absorber rods among them, keff values are smaller for ~0.2 (~0.1 for FLIP) than for arrangements without absorber rods almost regardless of the burnup. The effect of burnup is the smallest. For standard fuel elements, it is ~0.1 for almost all pitches and numbers of absorbers. For FLIP fuel, it is smaller for a factor of 3, but increases with the burnup for compact arrangements. Cooling time of fuel has just a minor effect on the keff of spent-fuel pool and can be neglected in spent-fuel pool design.