Search

Their review work summarizes various casting techniques to fabricate bronze alloys, mainly focusing on the microstructures, tensile properties, and tribological characteristics of Cu-Sn and Cu-Sn-Ti alloys [[19]]. 36676282 2 Li R., Xiao Z., Li Z., Meng X., Wang X. Work Hardening Behavior and Microstructure Evolution of a Cu-Ti-Cr-Mg Alloy during Room Temperature and Cryogenic Rolling. In order to obtain high-quality non-ferrous materials and/or components, we must fully understand the underlying process-structure-property relationship in the metallic materials such as Al alloys, Mg alloys, Ti alloys, Cu alloys, and etc. Huang et al. [[6]] elucidated the precipitation behavior of the T SB B sb (Al SB 7 sb Cu SB 4 sb Li) and S(Al SB 2 sb CuMg) phases of the 2195 Al-Cu-Li alloy at different homogenization temperatures. [Extracted from the article]

Radiative heat transfer between the different components of the nuclear reactor channel (fuel pins, pressure tube (PT) and calandria tube (CT)) is the predominant mechanism of heat transfer in a channel when coolant flow is ceased. In this paper, the thermal behavior of the channel of Indian PHWR that has thirty-seven fuel elements is analyzed for a postulated accident such as loss of coolant accident (LOCA) coupled with the failure of the emergency core cooling system. (ECCS). The analysis has been performed using experimental, numerical and analytical techniques. The experiment is performed under the pseudo-steady-state condition, and the temperature profiles for the different parts of a simulated channel are obtained. The numerical analysis of the above problem is carried out using the ANSYS® Fluent 19.0. The code calculation is compared with the exact solution which is obtained using the radiosity network method. There is a reasonably good agreement between these three types of analysis. The results from these analyzes show that the temperature gradient (along the radial direction) is developed in the fuel bundle simulator (FBS) from the center fuel element to the outer ring fuel element. Also, an insignificant circumferential temperature gradient is observed in the FBS, PT and CT. [ABSTRACT FROM AUTHOR]

This paper explains a preliminary study on the effect of oxidizing environment on the sagging deformation of Pressurized heavy water reactor coolant channel at extreme temperature. The coolant channel consists of two concentric tubes- calandria tube (outer tube) and pressure tube (inner tube). It is difficult to perform experiments with full-length coolant channels in the laboratory. Hence, a scaled-down concept is developed for scaling the coolant channels with a 1:3 scale. The various aspects of the geometric scaling of tubes are elaborated in this paper. The calculations for hydrogen generation are also presented analytically using existing and widely used correlations. The zirconium alloy of the calandria tube has reacted violently with steam, and the temperature escalation beyond 750°C produces significant chemical heat flux. The calandria tube has completely decayed during the experiment, and the channel is disassembled and fragmented into segments. Few segments are equal to one weight simulator length. Since the calandria tube cannot resist the deformation of the pressure tube due to loss in its structural strength, the pressure tube has undergone enormous unrestricted sagging before the channel failure. A maximum of 320 mm sag is reported at the mid-length of the channel. The scanning micrograph of the channel surface shows spalled oxide layers containing 45.89 weight percentage oxygen. [ABSTRACT FROM AUTHOR]

The sampling and measurement of airborne tritium is an essential component of workplace monitoring at heavy water handling facilities and nuclear reactors using heavy water as coolant and moderator. Tritium being an internal hazard, its workplace monitoring and assessment of internal exposure of workers is a regulatory requirement for the facility. The conventional tritium air sample collection methods are condensation, bubbling, and trapping with appropriate media such as dry ice, water, and desiccants, respectively. A novel method for rapid collection of moisture in the air for the estimation of tritium is presented and discussed in this article. It involves condensation of tritium oxide in the air using a commercially available thermoelectric cooling module which has removed uncertainty in the availability of dry ice or desiccant. The instrument is capable of collecting 2–3 ml of sample in 30 min at a relative humidity level of about 60% and temperature of about 25.5°C. The quantity of sample collected is sufficient for the estimation of tritium concentration in air. The Peltier module-based low-cost, simple, and reliable system has been successfully implemented for workplace tritium in air sampling at radiological facilities. [ABSTRACT FROM AUTHOR]

• CFD modelling undertaken for pressurized heavy water reactor fuel channel conditions. • Developed new wall boiling model for predicting CHF based on liquid film thickness. • Assessed model predictions against CHF tests for single-element configuration. This paper uses Eulerian multiphase computational fluid dynamics (MCFD) simulations to predict the critical heat flux in a simplified CANDU test geometry for typical pressurized heavy water reactor (PHWR) conditions. A diffused liquid film-based wall dryout model is proposed based on the physical description of the dryout phenomenon. Given the adopted ensemble-averaged two-phase framework, this proposed model can represent the film dryout physics but requires some calibration. Simulations are performed for three calibration cases at 9 MPa outlet pressures to determine the appropriate diffused liquid film layer thickness to accurately predict the experimental data for the test geometry. This calibrated layer thickness is then applied to a number of assessment cases for various axial heating profiles, mass fluxes, inlet subcooling levels, and outlet pressures. The accuracies of the assessment simulations are generally within ±11 % of the experimental measurements, with a few exceptions. Specifically, the errors for the cases at lower outlet pressures (6 MPa) and mass fluxes (≲ 2000 kg/m2-s) ranged from about −19 % to –33 %. The lower pressure cases indicate that a separate calibration may be required at low pressures, while the larger errors in the lower mass flux cases suggest that some improvement of the boiling closures is needed to account for conditions at high average outlet quality. [ABSTRACT FROM AUTHOR]

• Break opening time assumption in LBLOCA Large Break Loss of Coolant Accidents Analysis. • Demonstrating the efficiency of the second shutdown fast boron injection system during the accident analysis. • Dynamic Break Opening development for the analysis. In the frame of the licensing process of the Atucha II PHWR (pressurized heavy water reactor) the BEPU (Best-Estimate Plus Uncertainty) approach has been selected for issuing Chapter 15 of FSAR (Final Safety Analysis Report) dealing with accident analysis. In case of a LOCA (Loss of Coolant Accident) the assumption of instantaneous opening time is too conservative, especially for this kind of reactor with a positive coolant void reactivity coefficient. Analysis was performed to assess the ability of Atucha II nuclear power plant's fast boron-injection system to prevent overheating of the fuel elements in the event of a Beyond Design Basis Accident (BDBA). The goal of the Analysis was to develop a Dynamic Break Opening which was used as the input for the LOCA analyses. The results show that the fast boron injection can effectively shut down the reactor, thereby preventing the overheating of the fuel elements. [ABSTRACT FROM AUTHOR]

A dual sensitive, simple, and rapid membrane-based potentiometric sensor was developed to determine thorium (Th) ion in nuclear fuel samples which does not require any prior chemical separation of other matrix elements. Sensor was prepared by incorporating N,N,N',N',N",N" -hexa- n -octylnitrilotriacetamide (HONTA) in a polyvinyl chloride (PVC) matrix in presence of 2-nitrophenyl octyl ether (NPOE) and sodium tetraphenyl borate (NaTPB). The optimum membrane composition was found to be 14.1% HONTA, 3.2% NaTPB, 54.9% NPOE, 27.8% PVC, which showed a dual sensitive linear region in the calibration plot with sub-Nernstian and super-Nernstian slopes of 12.1 ± 0.4 mV/decade and 71.7 ± 3.5 mV/decade, respectively, with the corresponding linear dynamic ranges of 2.0 × 10−5 to 5.3 × 10−4 M and 3.8 ×10−3 to 1.4 × 10−2 M, respectively. The super-Nernstian sensitivity exhibited by the sensor increases with Th(IV) as well as HONTA concentrations in the membrane and can be thought to be due to micelle-mediated selective adsorption (MMSA) sensing of the metal ion. Solution speciation of Th(IV) in acetate buffer was determined by laser desorption/ionization mass spectrometry (LDI-MS) and Medusa software. Pristine and used membranes were characterized by Fourier transform infrared (FTIR), scanning electron microscopy–energy dispersive x–rays (SEM-EDS) and small angle neutron scattering (SAXS). Thorium was measured in an actual (Th–U) mixed solution and in a synthetic advance heavy water reactor (AHWR) fuel sample. The present study describes a new approach for the detection of Th(IV) ion through MMSA in an ion selective electrode (ISE). [Display omitted] • A highly sensitive and selective potentiometric sensor for Th estimation was developed. • Sensor exhibited dual sensitivity with sub and super-Nernstian slopes in calibration plot. • Aggregation of the ionophore has been observed from SAXS studies in the membrane. • Micelle mediated selective adsorption (MMSA) mechanism is proposed for sensing. • Sensor can determine thorium in mixed (Th 0.9 –U 0.1)O 2 samples without separation of other elements. [ABSTRACT FROM AUTHOR]

The connection between Zr–2.5%Nb pressure tubes and modified martensitic stainless steel-403 end fittings in Pressurized Heavy Water Reactors (PHWRs) at the rolled joint is a critical site marked by enhanced hydrogen diffusion and elevated tensile residual stresses. The application of circumferential stress above a threshold leads to the formation of radial hydrides. Formation of radial hydride leads to a drastic reduction in fracture toughness and increase in Delayed Hydride Cracking (DHC) velocity. A comprehensive understanding of residual stress, hydride orientation and radial hydride fraction in this region is essential for the assessment of pressure tube structural integrity. In this study, we employ 3D nonlinear finite element analysis to estimate the residual stresses near rolled joints for both 220 MWe and 700 MWe pressure tubes. Additionally, we pioneer an experimental investigation into hydride orientation near roll joints through internal pressurization of hydrided pressure tube and end fitting assemblies. The study delves into the influence of residual stresses due to rolling process on hydride orientation and radial hydride fraction, providing crucial insights for the evaluation of the structural integrity of pressure tubes. [Display omitted] • 3D nonlinear finite element simulation of Pressure Tube (PT)–End Fitting (EF) roll joining process. • Internal pressurization of PT–EF assembly. • Experimental correlation of the hydride orientation near rolled joints with computed residual stresses. [ABSTRACT FROM AUTHOR]

• Explored the factors influencing the cutting performance of plasma cutting on carbon steel plates of different thicknesses. • Provided data input for digital simulation of the decommissioning of the heavy water research reactor. • Successfully conducted a mock-up experiment on the plasma cutting of the heavy water research reactor outer shell under optimal cutting performance. • Offered guidance and personnel training for the actual decommissioning of the heavy water research reactor internals. The dismantling of internal components within the reactor is one of the primary technical activities in the reactor decommissioning project. Currently, the first nuclear reactor in China, heavy water research reactor (HWRR), is in the phase of decommissioning. By integrating the characteristics of the HWRR with both domestic and international cutting technology experience, plasma cutting has been identified as a suitable and commonly used technique for cutting the reactor internals. Considering that the decommissioning process of nuclear facilities is irreversible, it is essential to conduct mock-up experiments to acquire cutting performance parameters for the cutter prior to decommissioning of the reactor internals, which not only helps optimize the decommissioning plan but also provides data support for decommissioning practices. Based on the characteristics of the outer shell, within the HWRR, parameters such as cutting current, workpiece thickness, cutting speed, working angle, and nozzle standoff were selected for the plasma cutting experiments. Through orthogonal experiments, cutting current, workpiece thickness, and cutting speed were identified as key factors influencing the performance of plasma cutting. Based on the previous research, the maximum cutting speed and workpiece quality loss under various cutting conditions were determined through special experiments. Finally, using the outer shell mock-up of the HWRR as the experimental object and combining it with the decommissioning path provided by simulation, it was demonstrated that the optimal operating parameters obtained from the previous experiments can be successfully applied to practical outer shell cutting. [ABSTRACT FROM AUTHOR]

Recent developments have been made in a novel fuel for pressurised heavy water reactors: high burnup oxide fuels composed of a mixture of thorium and high assay low enriched uranium. Several claims are made regarding the performance of such fuels in terms of economics, waste and non-proliferation, among others. This article analyses this class of fuel using an infinite lattice study in SERPENT. First, a set of fuel compositions in the thorium/uranium enrichment design space that satisfy the high burnup target is produced. Then, a representative thorium-based fuel is compared to a uranium-only fuel with the same discharge burnup, as well as current standard fuel, in the context of the various performance claims. • A novel thorium-HALEU for use in pressurised heavy water reactors is considered. • Design space of thorium and uranium composition explored to achieve 55 MWd/kg burnup. • Claimed advantages in waste and non-proliferation are considered, as well as resource utilisation and reactor behaviour. [ABSTRACT FROM AUTHOR]

Tritium with different chemical forms have different effects on the human body. In tritium migration and conversion, soil plays a significant role. In this study, our samples are from the surrounding environment of the Qinshan Nuclear Power Plant, in addition, Qinshan Phase III Nuclear Power Plant is the only heavy water reactor nuclear power plant (CANDU type, 700 MWe) in China. Both tritiated water (HTO) and non-exchangeable organically bound tritium (NE-OBT) were measured at the surface soil (0–2 cm) at 8 sampling sites and their depth profiles in the subsurface 30 cm of soil in two sampling sites around Qinshan Nuclear Power Plant in 2015, 2016 and 2018. The distribution analysis showed the further away from the nuclear power plant, the lower activity concentration of HTO at the surface soil. Also, in the depth profile, the HTO activity concentration decreased as the depth increase. Through soil analysis at different depths of sampling points A and D, it is concluded that the activity concentration of NE-OBT is the largest in the surface soil of the selected sampling points, which is similar to the HTO of most sampling points. [ABSTRACT FROM AUTHOR]

When hydrogen isotopes are present in stoichiometric ratio of 2:1 with oxygen in inert gas such as helium, catalytic combustion is the most promising option to reduce the concentration of hydrogen isotopes and to minimize the associated safety hazards. This is the case, when deuterium and oxygen is formed during radiolysis of heavy water in moderator circuit of Pressurized Heavy Water Reactors (PHWR) and mixed with helium in the cover gas system. In order to design the catalytic reactor for recombination of deuterium and oxygen, the data on catalytic combustion kinetics under the similar process conditions are essential. However, the catalytic combustion data is generally reported for combustion of hydrogen in air, where oxygen is available in excess. The studies on the kinetics of catalytic combustion of deuterium where oxygen is present in stoichiometric ratio of 2:1 in an inert gas medium such as helium, is very scarce. In the present study, the multi-step reaction mechanism of catalytic combustion of hydrogen in presence of platinum catalyst is analyzed for the present process conditions. Based on the analysis, a simple rate expression is proposed for the present process conditions of catalytic combustion of deuterium. Further, 0.5% platinum catalyst is prepared on a stainless steel Dixon ring support with an objective to achieve better heat transfer characteristics and lesser reduction in the catalytic activity due to water adsorption. The kinetic data is generated using a differential packed bed reactor operating in a closed loop. The experiments were conducted at different temperatures using a stoichiometric mixture of deuterium and oxygen in helium. The rate constant for the above proposed model is estimated based on the experimental data at different temperatures. Further, the activation energy and frequency factor are determined and the activation energy for the present catalyst is found to be on the lower side in comparison to literature data. [Display omitted] • Studied the catalytic combustion of (2:1) mixture of D 2 with O 2 in He using Pt. • A simple rate expression is proposed based on detailed reaction mechanism. • Overall reaction rate is found to be limited by adsorption of oxygen. • The overall rate constant is estimated at different temperatures (35oC–145 °C). • The apparent activation energy and frequency factor were estimated. [ABSTRACT FROM AUTHOR]

Indian pressurised heavy water reactors use Incoloy 800 for steam generators. The pH is maintained alkaline to check corrosion in secondary circuit by the addition of Ethanolamine (ETA). Effect of octadecylamine (ODA) for further reduction of corrosion is investigated. Incoloy was exposed to both steam and liquid phases at 200°C in an autoclave for 7 days. Concentration of ETA was kept constant at 6 ppm and that of ODA was varied from 6 ppm to 20 ppm. The film grown in liquid phase was thinner and enriched with Cr, while the film grown in steam phase had larger particles and mixed-ferrite in nature. Electrochemical Impedance spectroscopy studies suggested the presence of a barrier film under both liquid and steam phase oxidation. The addition of ODA was comparatively more beneficial in liquid phase. The study showed that ODA is a promising amine to inhibit the corrosion of Incoloy 800 in secondary water chemistry conditions. [ABSTRACT FROM AUTHOR]

The majority of data reported in open literature on hydride embrittlement of Zr alloys is based on hydrogen not deuterium which evolves during aqueous corrosion of in-core components in pressurized heavy water reactor operations. It is assumed that hydrogen and deuterium have similar effects on tensile properties and fracture toughness of Zr alloys. The thermal and physical properties as well as stress free transformation strain of hydrides and deuterides are different. Because of these differences, hydrogen and deuterium may affect tensile and fracture properties of Zr alloys to a different extent. The hydrogen isotope effects on tensile and fracture properties were investigated using samples machined from Zr–2.5Nb pressure tube (PT) material charged with hydrogen between 30 and 200 wppm and tests were carried out between 25 and 300 °C. The tensile properties and fracture toughness of the hydrogen charged Zr–2.5Nb PT material were compared to the Zr–2.5Nb PT material charged with deuterium. The observed effect of hydrogen isotopes on tensile and fracture toughness was rationalized in terms of stress free transformation strain and the temperature dependence of the fracture toughness of the hydrided material was correlated with the presence of axial splits on fracture surfaces. The effect of hydrogen isotope on tensile and fracture properties was investigated for the Zr–2.5Nb pressure tube (PT) material charged with hydrogen between 30 and 200 wppm and in the temperature range of 25 and 300 °C and compared with the corresponding values determined for the Zr–2.5Nb PT material charged with deuterium. It was observed that for a given concentration up to 200 wppm, hydrogen and deuterium charged Zr–2.5Nb PT material have practically same tensile and fracture properties at all test temperatures between 25 and 300 °C indicating no effect of nature of hydrogen isotope on tensile and fracture properties of Zr–2.5Nb PT material. [ABSTRACT FROM AUTHOR]

This set out " to formulate, where possible, As noted already in the paper, the UK ALARP approach commonly shared safety concepts", noting that in general the and the post-TMI USNRC approach recognised both the use concepts included were not new but represented the best (then) of risk informed decision making and the use of qualitative current safety philosophy. We will Probabilistic safety assessment (PSA) provides one of the therefore discuss both the UK approach and the USNRC and how inputs into the ALARP process and has been used in the UK as RIDM has developed in each. an input into both design and assessment of nuclear power plants since the early 1970s. However much care is taken in developing the 0 [15] USNRC, "An approach for using probabilistic risk assessment in risk-informed decisions on plant specific changes to the licensing basis., Regulatory Guide 1.174", USNRC, 2011. The Trials and unlikely events, which allowed conservative performance The MCA expanded to become a set of stylised and very criteria to be defined for the key systems and therefore defined Tribulations analysed to allow for uncertainties. [Extracted from the article]

A conductivity-based technique is developed for the determination of Gd3+ in the heavy water moderators of pressurized heavy water reactors (PHWRs). The method involves monitoring extremely small shifts in conductivity, in the order of few nS/cm, due to the continuous addition of a suitable complexing agent to Gd3+ in aqueous medium. The resulting plot gives two distinct regions with vastly differing slopes. Two multidentate ligands, ethylenediaminetetraacetic acid (EDTA) and diethylenetriaminepentaaceticacid (DTPA), as complexing agents are compared. A high performing conductivity detector based on a new class of sensors called pulsating sensors that works entirely in the digital domain is deployed to monitor the conductivity shifts. Titration plots are studied in both H2O and D2O, and the observed difference between the plots in the two matrices is discussed in detail. Boron did not interfere in the analysis. The method was validated using the UV–vis spectrophotometric technique. The method is sensitive and rapid, as each analysis takes 3 min. The limit of detection in H2O and D2O are 1.27×10−7 mol/L and 5.1×10−7 mol/L, respectively. The precision in analysis lies between 1.9% and 5.3%. This method has important application in the nuclear industry for the routine analysis of gadolinium. [ABSTRACT FROM AUTHOR]

A two-step ion exchange process has been developed for decontamination of used pressure tube zircaloy, from pressurized heavy water reactors, with respect to 125Sb and 60Co. The steps are: anion exchange separation of Sb (III) and cation exchange separation of Co (II); from Zr (IV). Our earlier results on distribution coefficients, isotherms and kinetic separation of Co were the erecting pillars of the present integrated separation process. Oxidation state of Sb in the feed solution, need for reduction of Sb (V) to Sb (III), the nature and amount of reagents required were investigated. Effect of Zr concentration on uptake of Sb and Co, resin capacity, requirement, column dimensions, and flow rate was evaluated. [ABSTRACT FROM AUTHOR]

The physics, experimental setup, and analysis of results involved in Walther Bothe and Peter Jensen's mistaken 1941 measurement of the capture cross section for thermal neutrons by carbon are examined. Their experiment, while well‐conceived and executed, was corrupted by the presence of unappreciated impurities. This erroneous measurement was crucial as it prompted a decision by German military administrators to abandon graphite as a possible moderator for a nuclear pile in favor of heavy water, a decision which set their program back irreparably. In early 1941, Walther Bothe measured the cross‐section for capture of thermal neutrons by graphite as part of the wartime German nuclear program. The graphite contained impurities, which resulted in a large cross‐section and its rejection as a moderating material. This paper analyses the experiment and speculates on the nature of the impurities that may have been involved. [ABSTRACT FROM AUTHOR]

Uranium is the basic raw material for nuclear energy and is quite highly regarded. Developing a safe supply of uranium is important for safeguarding sustainable nuclear development. The purpose of this study is to evaluate the sustainability of uranium development in China based on dynamic system modeling combined with GAN (Generative Adversarial Network) analysis. We considered eight essential indicators and 42 sub-indicators as part of a detailed quantitative description, and then developed a framework to evaluate and rank China-specific sustainability in light of the quantitative performance of five options for fuel cycle transition scenarios. We began by using KMO sample measurements and the Bartlett Test of Sphericity to determine the suitability of factor analysis and the fitness of the corrected model map and observation data. We then analyzed the roles of different representatives of the decision makers and their impacts on the overall ranking by applying GAN methods from a weighted perspective. Five transition scenarios identified are 1) Pressurized Heavy Water Reactors, 2) Mixed Light Water Reactor + Fast Reactor, 3) Mixed LWR+FR fuel cycle scheme with heterogeneous irradiation, 4) Mixed Pressurized Water Reactor + FR fuel cycle scheme with plutonium recycled directly and repeatedly, and 5) Sodium-cooled fast breeder reactor power plant. The results showed that scenario 1 is the most unsustainable and highly confrontational scenario with a high demand for uranium resources, the lowest sustainability and a high level of antagonism among departments. On the other hand, Scenario 5 requires more advanced technology but exhibits less antagonism among the departments, and thus it largely satisfies the basic requirements for uranium sustainability and low levels of antagonism. In this paper, a safety assessment index system for the uranium supply is computed using a GAN framework. This system plays a crucial role in the sustainable supply and development of uranium, and provides flexibility for coping with the evolution and inherent uncertainties of the necessary technological developments. [ABSTRACT FROM AUTHOR]

Decommissioning of nuclear power plants is a multistage process involving complex operations like radiological characterization, decontamination and dismantling of plant equipment, demolition of structures, and processing and disposal of waste. Radioactive effluents released into the environment may result in exposure of population through various exposure pathways. The present study estimates the public dose due to atmospheric discharge of important radionuclides during proposed decommissioning activities of Indian Pressurized Heavy Water Reactors. This study shows that major dose contributing radionuclides are 60Co followed by 94Nb, 134Cs, 154Eu, 152Eu, 133Ba, 99Tc, 93Mo and 41Ca. It is found that infant dose is higher than adult dose and major fraction of total dose (~98%) is through ground shine and ingestion; other pathways such as inhalation and plume shine contribute only a small fraction. This study will be helpful in carrying out radiological impact assessment for decommissioning operations which is an important regulatory requirement. [ABSTRACT FROM AUTHOR]

Large nuclear reactors exhibit spatial oscillations in neutron flux distribution on account of xenon reactivity feedback. These oscillations, if not suppressed, can pose nonuniform spatial distribution of power and raise chances of fuel failure. Thus, it is necessary to design a control strategy to regulate these spatial oscillations. This article presents the design of an integral sliding mode control (ISMC) for spatial power control of the advanced heavy water reactor (AHWR). The AHWR model considered here has 18 outputs as well as 4 inputs and is characterized by 90 state variables. This nonlinear model is linearized around an operating point and an optimal controller is designed. To ensure the robustness of the system under external disturbances, ISMC is designed. The finite-time convergence of the controller is provided by the Lyapunov approach. A comparative simulation study is presented to show the efficacy and robustness of the proposed controller using the nonlinear model of AHWR. [ABSTRACT FROM AUTHOR]

Purpose: Advanced heavy water reactor (AHWR) is a pressure tube type of heavy water reactor. It eliminates high-pressure heavy water coolant resulting in a reduction of heavy water leakage losses and eliminating heavy water recovery system. It recovers the heat generated in the moderator for feed water heating. However, it requires a satisfactory technological response to develop an effective controller that attains the challenges of the very high-level safety system. Hence, they require application-specific improvement for better controlling performance. Design/methodology/approach: The purpose of this study intends to propose a system for controlling state vectors v1 and v2and in AHWR using Grey Wolf second-order sliding mode control (GW-SoSMC) technique. The main aim of the paper is to minimize the errors between the predicted and desired azimuthal angles of the system. With this proposed method, it is possible to mitigate both the chattering phenomenon and controlling performance of AHWR system. It implements a SoSMC controller based on GWO algorithm for the purpose of controlling the state vectors in the AHWR system. It aims to accomplish a controller for improving the performance of the AHWR system. Findings: Through the performance analysis, the efficiency of the proposed GW-SoSMC technique was verified by comparing it with various conventional algorithms, such as GW-SMC, FF-SoSMC, ABC-SoSMC, GS-SoSMC and GA-SoSMC. From the analysis, it was obtained that the implemented GW-SoSMC technique was 65.3 per cent superior to GW-SMC, 65.32 per cent superior to both FF-SoSMC and 65 per cent superior to ABC-SoSMC, 65.8 per cent superior to the GS-SoSMC and 58 per cent superior to the GA-SoSMC methods. Thus, the effectiveness of the proposed method in controlling the state vectors in AHWR was obtained. Originality/value: This paper presents a technique for controlling the state vectors in the AHWR system using GWO algorithm. This is the first work that uses GWO-based optimization for controlling state vectors in the AHWR system. [ABSTRACT FROM AUTHOR]

The dismantling of structural objects during the decommissioning of nuclear facilities needs radioactive source characterisation for the planning of decommissioning strategies in compliance with the ALARA (as low as reasonably achievable) principle. The sources may arise from neutron activation of the structural components in the reactor core as well as contamination due to the radioactive release from the fuel occurred during normal operation or unplanned events in a nuclear power plant (NPP). In a pressurised heavy water reactor (PHWR), various in-core components are predominantly made of either zircaloy–2 or 4. The nuclides present as impurities in the zircaloy, playing a crucial role in the activity inventory due to neutron activation of those nuclides, which in turn determine the external gamma dose rate. The activity of the activation products depend on the neutron flux seen by the component, duration of irradiation and cooling period, half-lives of the daughter products and the amount of the impurities present in the structural components. To illustrate this, a guide tube made up of zircaloy–4 has been considered. A guide tube assembly is a part of the primary shut down system (PSS) which guides the movement of absorber elements in the upward and downward direction in the calandria. This study has identified and quantified the activity inventory in a guide tube at the end of the operation of the reactor using the ORIGEN2 code, and then estimated the associated external gamma dose rate using the FLUKA Monte Carlo code. The findings will help the management of radioactive waste, cost optimisation and collective dose budgeting during the decommissioning stage of a typical PHWR. [ABSTRACT FROM AUTHOR]

The super multi-functional calculation program for nuclear design and safety evaluation is a general, intelligent, accurate and precise simulation software system for the nuclear design and safety evaluations. The heavy water reactor has a much stronger moderation power and much longer diffusion length of the thermalized neutrons. The paper intends to show the verification and validation of SuperMC3.2 with a heavy-water-moderated lattice named the deuterium critical assembly which is very similar to the Canada Deuterium Uranium type reactor and selected from the international reactor physics experiment evaluation project. The calculation results were compared with the reference calculated results and the experimental data from International Reactor Physics Experiment Evaluation Project. The final obtained results proved the accuracy, convenience and universality of SuperMC, and primarily verified the applicability of SuperMC in nuclear analysis of heavy water reactor. [ABSTRACT FROM AUTHOR]

The Zr-2.5 %Nb is used as the Pressure Tube (PT) material in Pressurized Heavy Water Reactors (PHWRs) in both cold worked and stress relieved as well as heat treated condition. For the development of heat-treated pressure tube, process parameters need to be tuned to achieve optimum microstructure and texture. In microstructural aspect, prior β grain size is a very important factor in influencing the property of the material. Hence, the prior β grain growth kinetics of the Zr-2.5 %Nb alloy has been investigated by subjecting the material to different heat treatments. Different heat treatment temperatures have been chosen from (α+β) and β phase region. The β transus temperature was measured through dilatometry. Specimens were soaked for different time and the grain size were measured using optical metallography. The evolution of grain size with soaking temperature and time were analyzed to study the grain growth kinetics. [ABSTRACT FROM AUTHOR]

• Thermal-hydraulic analyses of the exposed reactor-scale core geometry for row-wise exposures of fuel channels. • Terminated at the onset of fuel channel disassembly. • Convection heat transfer is dominant for initial exposures of the rows of fuel channels. • Advanced stages of fuel channel exposures governed by thermal radiation. • Fuel channel disassembly is restricted to central regions of the exposed core. A two-dimensional steady-state thermal-hydraulic analyses of the exposed core with varying levels of exposure is carried out for heat transfer assessment. Low-frequency postulated severe accident without mitigating actions leads to exposure of fuel channels in Pressurised Heavy Water Reactors (PHWRs). Such exposure of fuel channels leads to their heat-up which is governed by interdependent heat transfer mechanisms such as radiation, convection and conduction heat transfer. In order to understand such a complex behaviour as well as to plan the severe accident management guidelines, numerical analyses of the exposed core are required. In this paper, one-of-a-kind thermal-hydraulic analyses of the exposed reactor-scale core geometry are presented for row-wise exposures of fuel channels. These numerical analyses are terminated at the onset of fuel channel disassembly i.e., loss of geometrical intactness of the exposed core. The interesting interdependency of convection and radiation heat transfer on fuel channel heat-up is brought out in this study. The dominance of convection heat transfer is observed for initial exposures of the fuel channels. However, for the advanced stages of fuel channel exposures, the heat transfer is observed to be dominated by radiation heat transfer. Intriguingly, it is observed that the convective cooling of exposed fuel channels by steam flow is counter-acted by exothermic heat generation arising from Zircaloy and steam interaction. Moreover, it is interesting to note that the disassembly of fuel channels is restricted to central regions of the exposed core. [ABSTRACT FROM AUTHOR]

• The cooling capacity of moderator decreases with moderator depletion. • The highest temperature reached in the PT for 3500 W/m2 heat input was 1684.3℃. • The highest temperature reached in the CT for 3500 W/m2 heat input was 254.51℃. • Moderator acted as an efficient heat sink up to 20 mm moderator depletion at 3500 W/m2 heat input flux. A steady-state heat transfer analysis is conducted on a completely voided concentric nuclear channel within an Indian pressurized heavy water reactor (IPHWR) to evaluate the temperature distribution in both the pressure tube (PT) and calandria tube (CT), taking into account the depleting moderator condition. The simulations encompass a range of five distinct moderator levels i.e., 10–30 mm at an interval of 5 mm by varying the input heat flux in the fuel bundle at values of 980 W/m2, 1500 W/m2, 2500 W/m2, and 3500 W/m2. The results showed that as the moderator depletes, the temperature difference from top to bottom exhibits an increasing trend in both the pressure tube (PT) and the calandria tube (CT). The maximum top to bottom temperature deviation in the CT was 83.21 %, and in the PT was 0.34 %, at an input heat flux of 2500 W/m2 with 30 mm moderator depletion. The CT is observed to attain a boiling temperature of water at 30 mm moderator depletion for the heat flux of 980 W/m2. Further, for higher heat inputs i.e., 2500 W/m2 and 3500 W/m2, CT attains the temperature more than the boiling point of moderator at initial depletion level which means that the boiling of the moderator starts form the initial depletion level. This scenario may further lead into moderator voiding which may results into the meltdown condition. These results will be helpful to determine the conditions where the temperature rise becomes uncontrollable, and to access the threshold condition for meltdown. [ABSTRACT FROM AUTHOR]

In view of the vast resources of thorium in India, thorium-based fuel cycle has been envisaged as one of the most promising options to meet sustainable energy requirements. In this direction, ThO 2 −233UO 2 and ThO 2 −UO 2 (low-enriched) mixed oxide fuels with < 30% UO 2 have been proposed for the Advanced Heavy Water Reactor. In the present work, phase relations in the Th 1-x U x O 2 + δ (x ≤ 0.3) system under reducing and oxidizing conditions have been investigated by means of X-ray diffraction and Raman spectroscopic studies. Diffraction studies confirmed the formation of single-phase fluorite-type solid solutions; however, the samples synthesized in air exhibited slightly lower lattice parameters than those of their reduced counterparts, attributed to oxidation of U4+. Raman spectroscopic studies indicate the formation of various defects upon uranium substitution in parent ThO 2 , depending on synthesis conditions. It also highlights the existence of more disorder in the air-synthesized samples as compared to their reduced counterparts, plausibly due to the presence of a variety of oxidation states of uranium and associated oxygen anions interstitials. Thermal annealing of the solid solutions under air at T ≥ 1873 K results in heavy uranium evaporation loss, which has been established by XRD and XRF studies. [ABSTRACT FROM AUTHOR]

In laboratories dealing with radioactive samples it is important to minimize both the sample size and also the associated waste generated in an analysis. To meet this objective a rapid conductometric titration technique is developed to determine boron in the moderators of Pressurized Heavy Water Reactors (PHWR's). Using this novel PC interfaced titration facility a minimum tenfold reduction in sample size is achieved compared to conventional conductometric titration. Determination of boron is based on the conversion of extremely weak boric acid to better conducting boron mannitol complex and titrating the complex against NaOH. Various parameters affecting the analysis, when moving from large to small sample size, are analyzed and optimized. The technique is primarily proposed for the assay of boron (≥0.5 ppm) during reactor startup. Each analysis requires less than 10 min. The detection limit is 0.5 ppm and the precision obtained at this level is 4.6% RSD. The technique is a good alternative to less sensitive carminic acid based spectrophotometric method. [ABSTRACT FROM AUTHOR]

• The one-step explicit, implicit and the midpoint integration schemes are examined for numerical stability. • The coupled dynamics of an Advanced Heavy Water Reactor is adopted as the case study. • A bifurcation framework is resorted for the stability analysis. • Super-stability features are figured out for the implicit schemes while the explicit method is quite prone to instability. The tendency of certain numerical discretization methods towards super-stability or instability is investigated in this work. To this end, the coupled ordinary differential equations which govern the kinetic behavior of a class of Advanced Heavy Water nuclear Reactors (AHWR) are taken into account. Single step explicit and implicit methods and the multi-step central difference scheme, also known as the leapfrog or the midpoint method, have been applied to these coupled ODEs. Pertaining stability regions have been delineated in the parametric plane characterized by the pair of inherent thermal feedback coefficients (α v , α f) present in the model. To this effect, the bifurcation framework has been employed and the discretized schemes were compared to the original (continuous) system. Results pointed out that the implicit schemes enjoy a super-stable behavior for large enough time-steps while characterizing a broader stability region. The explicit scheme on the other hand is quite susceptible to an unstable dynamic regime. Besides, the super-stable behavior is quite more pronounced in the midpoint rule owing basically to the higher order accuracy thereof. [ABSTRACT FROM AUTHOR]

• A novel concept of molten salt reactor moderated by heavy water (HW-MSR) is proposed. • HW-MSR inherits the positive properties of MSR and heavy water reactor. • HW-MSR can address the problems of depleted graphite management and positive temperature coefficient existed in MSR. • BR and doubling time of HW-MSR are respective 1.073 and 12 years at the reprocessing cycle of 10 days. The Heavy Water moderated Molten Salt Reactor (HW-MSR) is a novel concept of a thermal heterogeneous nuclear reactor and pursues for Thorium-Uranium (Th-U) breeding. It adopts heavy water rather than graphite as moderator while employs the same liquid fuel of molten salt reactor (MSR). Thus this new concept inherits the positive properties of MSR (feasible for online reprocessing) and heavy water reactor (high neutron moderating ratio). Furthermore it can address the problems of depleted graphite management and positive temperature coefficient due to neutron spectral shift propitious to 233U fission existed in Molten Salt Breeding Reactor (MSBR). HW-MSR consists of the primary loop system, intermediate cooling loop, energy conversion system, external cooling system and online reprocessing system. Since there is a huge temperature difference between molten salt (higher than 600 °C) and heavy water (lower than 100 °C to keep high density) in the core, effectively preventing the heat transfer between them is one main challenge. To address this problem, a thermal insulator made of Yttria Stabilized Zirconia (YSZ) is applied and analyzed. The calculation results show that both the fuel salt and moderator outlet temperature across the core can satisfy the design requirements. Similar to a traditional MSR, the HW-MSR also implements an online reprocessing system which can online extract and recycle useful fuel during operation, providing a feasible approach for Th-U fuel cycle. In addition, because of the outstanding neutron performance of heavy water moderator, a small initial 233U loading mass is required for HW-MSR, and its breeding ratio can achieve 1.073, corresponding to a doubling time of 12 years at the reprocessing cycle of 10 days. Compared with traditional light water reactor (LWR) whose transuranium (TRU) production is 69 kg/GWy, HW-MSR produces only 0.51 kg/GWy. This much lower radiotoxicity of nuclear waste in the HW-MSR would significantly enhance the sustainability of nuclear energy. [ABSTRACT FROM AUTHOR]

Evolutionary algorithms play an important role for solving various optimization problems related to fuel management in reactor physics like core loading pattern optimization (LPO) or refueling. In general, all algorithms make a sample of solution candidates and evaluate the fitness of all candidates, and then the candidates with better fitness value are used to generate the next sample of solution candidates. In optimization algorithms, internal parameters [like population size, weighting factor in estimation of distribution algorithm (EDA) and population size, cross-over rate, etc., in Genetic Algorithm (GA)] have a stiffness problem as the value of these parameters is fixed at the first generation and is not being changed subsequently. However, the flexibility of changing the value of even one internal parameter during the generations will make the algorithm more efficient. In this paper we propose that fuzzy logics can be used in an innovative way to eliminate the stiffness problem related to internal parameters in evolutionary algorithms. As a test case, EDA for initial core LPO of the advanced heavy water reactor is chosen, and the use of fuzzy logics has shown a significant improvement in the algorithm's performance. The appropriate value of weighting factor α in EDA has been predicted using fuzzy logics in each generation, and this has resulted in efficiency improvement of the algorithm. The improved methodology is expected to give better performance with other optimization algorithms, such as the GA or Ant Colony Optimization Algorithm, etc. [ABSTRACT FROM AUTHOR]

• Enhanced neutron multiplication in subcritical blankets of fusion-fission hybrids. • The hybrids as multifunctional systems acting also for expanded tritium production. • Nuclear performances and restrictions of the water and gas cooled blankets for DEMO-FNS. While the future fusion power plants (FPP) will consume and breed tritium for their own use, an essential amounts of tritium will be needed additionally from external sources for the FPPs initial start-ups in the commissioning periods and after technical interruptions that result in plasma operation shutdowns and the following a year-scale recovery time. Up-to-date predictive estimates of the start-up T-inventories for a single FPP give the values comparable with the commercial tritium resources currently available in the world nuclear industry or even exceed those. The neutron rich fusion-fission hybrid reactors (FFHR) with subcritical blankets controlled by a fusion neutron source are generally considered as possible producers of several product types as nuclear fuel for fission systems, energy and neutrons for waste burning. In this study, the hybrids are regarded as multifunctional systems acting also as tritium breeders. Some intrinsic properties and distinctive features of such systems are noted with emphasis on the nuclear performance variation during the operation periods. It is shown that a tritium surplus of a few kilograms per year may be produced even in a FFHR with a modest (˜50–100 MW) fusion driven neutron source during 3–4 operation years. Such productivity exceeds essentially the tritium production rate of about 0.2 kg/year at a typical commercial CANDU heavy water nuclear reactor. [ABSTRACT FROM AUTHOR]

The paper deals with the safety evaluation and the embedded licensing process of the Atucha-II 800 Mwe Pressurized Heavy Water Reactor (PHWR) in Argentina. Atucha-II was designed by Kraftwerk Union (KWU) during the 60's, the related construction was started/stopped in the 90's and restarted on 2006, and was connected to the electrical grid in 2014. Because of market policies, the KWU designer could not be directly involved in the licensing process during the first decade of the 2000 millennium: a licensing suited safety evaluation was performed by the Nucleoeléctrica Argentina (NA-SA), utility owner of the Atucha-II, with support of external experts group. A designer-independent assessment was performed having access to the installed systems and components other than the relevant design documents. Large core size (related to Pressurized Water Reactor – PWR – producing the same thermal power), presence in the core of natural uranium and heavy water fluids, i.e. the coolant and the moderator driven by two circulation loops with different average temperatures, characterize the system design. In those conditions, during the early phase of a depressurization transient, the 'hot' coolant vaporizes and the colder moderator remains in the liquid phase: a positive void coefficient is created. The relevance of the Large Break Loss of Coolant Accident (LBLOCA) in safety assessment is discussed with emphasis given to the system design features, the approach pursued in the analysis and the key results. Break opening time and time of occurrence of the safety boron injection affect, during the early period of the transient, the propagation of (negative) pressure wave, the fluid flashing, the heat transfer and the neutron flux: a fission power excursion is expected to occur. The analysis of a complex three-dimensional situation, considering the unavoidable uncertainties associated with the computation, demonstrates that safety limits are preserved. [ABSTRACT FROM AUTHOR]

A rapid high performance liquid chromatography (HPLC) and thermal ionisation mass spectrometric (TIMS) methods have been developed for the separation and estimation of fission product elements molybdenum and lanthanum for the burn-up measurements on the dissolver solution of Indian pressurised heavy water reactor (PHWR) spent fuel. Reverse phase chromatography method was developed to separate molybdenum from dissolver solution using mandelic acid as mobile phase and a dynamic ion exchange chromatography technique was used for the separation of lanthanum as well as neodymium from a dissolver solution. Sample loading methods which resulted in enhanced ionisation efficiency have been developed for the TIMS analysis of HPLC separated molybdenum and lanthanum fractions. Ascorbic acid mixed with silicic acid in HCl medium was used for loading the molybdenum on to a rhenium filament to obtain stable and intense ion beam. A novel sample loading method for lanthanum in which a mixture of graphite + boric acid (H3BO3) + silica gel was employed to achieve enhanced and steady ion beam formation of LaO+. Concentrations of species of interest were determined employing suitable spikes by isotope dilution mass spectrometry (IDMS) method. The developed methods were adopted for PHWR dissolver solution to establish molybdenum and lanthanum as alternate burn-up monitors. The burn-up data obtained were compared with the well established method of neodymium as fission product monitor. This is a first study of its kind where the data obtained by using molybdenum and lanthanum as fission product monitors were compared with that obtained by Nd-148 method. [ABSTRACT FROM AUTHOR]

Signals from sensors placed at different locations are used for control and protection purposes in a nuclear reactor. It also requires in-service fault detection and isolation (FDI) for its safe operation. The sensor signals are generally a superimposition of low-frequency components representative of true values of the variables being monitored, occasional high-frequency periodical oscillations due to disturbances and faults, and sensor faults. Techniques like principal component analysis (PCA) can be used for FDI; however, it would be more meaningful if the FDI technique can also help for predictive maintenance of reactor internals through vibration spectra. To address these issues, a multiscale PCA, integrating wavelet transform with PCA and aiming to reduce the modeling cost by using only a fewer scales that contribute to the monitoring, has been proposed in this paper for online FDI of an advanced heavy water reactor (AHWR). A new mathematical formulation of the generalized likelihood ratio test for its use with wavelet approximation coefficients has also been proposed for better sensor-FDI outcomes. The proposed approach detects and isolates sensor faults and process faults using the signals from neutron detectors. Efficacy of the proposed technique is established on the simulated ex-core ion chamber data of AHWR considering different scenarios that involve localized frequency contents representative of process faults, slowly developing (incipient) sensor faults, and the simultaneous presence of two or more of these scenarios. Simulation results validate the effectiveness of the proposed scheme for online FDI in the reactor. [ABSTRACT FROM AUTHOR]

A simple, simultaneously multi-elemental, ICP-AES based method was optimized to characterize nuclear high level waste solutions from different origin: research reactors, pressurized heavy water reactor, thoria containing waste and 137Cs rich solutions. It involves identification of interference free analytical lines with the evaluation of their analytical performance including sensitivity, detection limit etc. To compare the results of ICP-AES system, ED-XRF technique has been employed and the results compared well indicating the efficacy of the optimized procedure. For analytes like Fe, Cr, Na, Mn and Ca the detection limits are in the range of 0.001–0.009 mg/L while for other analytes (except U and Th) the detection limits are almost on the order of magnitude poorer. For all the analytes the linear dynamic range exists from the detection limit up to 500 mg/L. [ABSTRACT FROM AUTHOR]

Purification of heavy water used as the moderator and coolant of pressurized heavy water nuclear reactors (PHWRs) is a vital step in ensuring its adequate isotopic content that maintains neutron economy in the reactor core. This process is carried out in a packed vacuum distillation column which requires condensers and reboilers as essential heat transfer components. A vertical thermosiphon reboiler is used to minimize operating costs of the distillation system by eliminating the fluid recirculation pump and improving system reliability and safety by minimizing leakages and heavy water losses. This work presents a simple, direct methodology for optimizing the design and operation of such a reboiler using codes developed in-house and based on heuristics and two-phase heat and momentum transfer correlations from literature. The methodology is general and can be applied to any application requiring a thermosiphon reboiler. Parametric analysis and cost calculations have also been performed along with exergy analysis of reboiler performance. [ABSTRACT FROM AUTHOR]

Supercritical water is uniquely a green medium for diverse applications because of its changing nature from polar to non-polar. Owing to this property, it is being considered for heavy oil upgradation since it dissolves both organics (oil) and hydrogen while inorganics behave conversely. However, because of the high pressure and temperature (22.1 MPa, 374 °C), corrosive environment (chlorides, sulfides and salt deposition) and stresses involved, there are serious concerns encountered while utilizing supercritical water in reactors. These include change in the component-material microstructure due to hydrogen ingress, sulfide stress corrosion cracking and salt deposition leading to pitting and de-alloying. Various alloys such as ferritic–martensitic steels, austenitic stainless steels, Ti-, Ni- and Zr-based alloys have been used, while new alloys and materials are continuously being investigated to considerably help abate these problems and ultimately improve the life of reactors. Despite significant past efforts in material development, reactors still suffer from these problems and challenges. This review assesses materials selection, the current progress in material development as well as their potentials in ameliorating reactors resistance to oxidation, pitting, embrittlement, etc. This study aims to improve understanding of material selection for supercritical water reactors based on the corrosive environment of the reactor and hence help engineers to make insightful decisions in selecting material for the specific corrosive environment. Schematic illustration of materials susceptibility in supercritical water reactor [ABSTRACT FROM AUTHOR]

Highlights • The pseudo state temperature of 900 °C−1000 °C was obtained in the fuel bundle for 1% decay power. • Insignificant circumferential temperature distribution was found in the fuel pins. • 85% Decay heat was removed by moderator during large break LOCA. Abstract Postulated accidents like large break LOCA leads to expulsion of coolant in the primary heat transport system thus voiding of reactor core initially. However, the reactor is shut down at the onset of LOCA along with coolant injection from ECCS to remove the decay heat of 2–3% of nominal power. Further postulation of failure of ECCS leads to rapid increase in the temperature of fuel pins and the coolant channel as well. The moderator present around coolant channels limits the rise in temperature. The assessment of temperature distribution in the fuel pins of the bundles of a channel under high temperature is quite important from hydrogen generation and fission product release point of view. A pseudo steady state experiment has been carried out to obtain the temperature distribution of different components of a simulated coolant channel for large capacity Indian pressurized heavy water reactor. The experiment simulates a postulated LOCA with Loss of ECCS scenario for the coolant channel. The experimental results shows that for 1% decay power the simulated fuel pins attained a maximum pseudo steady state temperature of 900 °C–1000 °C, thus establishing moderator as a heat sink. An insignificant circumferential temperature variation in the channel components is observed which indicates weak effect on natural convective heat transfer within coolant channel. Estimation shows that around 85 percent of total decay heat provided through electrical power is removed by the moderator. [ABSTRACT FROM AUTHOR]

• The overall plant system related to the water-cooling system has been developed for JA DEMO. • Management of tritium concentration in the primary cooling system is clarified for JA DEMO. • Tritium permeation from the primary to the secondary cooling system was found to be less than the Japanese PWR restriction value. • A pressure suppression pool is effective to suppress the tritium release to the environment at the ex-vessel LOCA. • The system diagram of water coolant and tritium was also delineated for JA DEMO. The conceptual design of Japan's fusion demonstration plant (JA DEMO) is now being developed. In this paper, an overall plant system concept related to tritium handling in the water-cooling system is developed to give a concrete shape to the present JA DEMO concept as an electric power plant. The basic condition of tritium permeation from the in-vessel components to the primary cooling system is evaluated to be 5.7 g-T/day. The tritium concentration of the primary coolant is assumed to be 1 TBq/kg similar to the heavy water reactor condition. The capacity of the water detritiation system (WDS) is assessed, and the bypass feed water from the primary cooling loop is evaluated to be 94 kg/h under the tritium extraction efficiency of 0.96. Based on those specific parameters, the existing WDS in the heavy water reactor is found to be applicable to that of JA DEMO. Configuration of the primary heat transfer system (PHTS) is also discussed. Based on the heavy water reactor experience, tritium permeation through a steam generator (SG) to the secondary cooling system in PHTS is evaluated at 11.77TBq/year/loop (318 Ci/year/loop), which is found to be less than the restricted amount of tritium disposal for a pressurized water reactor in Japan. The key effect of the heavy water reactor experience is reduction of tritium permeation by oxide layer formed on SG pipes. Finally, a confinement concept of tritium release from PHTS is discussed under the condition of ex-vessel loss of coolant accident (LOCA). A pressure suppression system is installed to prevent the upper tokamak hall from pressurizing at the ex-vessel LOCA, and the tritium leakage from the upper tokamak hall is consequently restrained. The resultant early public dose at the plant site boundary can be reduced to 1.8 mSv, which is negligibly smaller than 100 mSv of the no-evacuation limit recommended by IAEA. [ABSTRACT FROM AUTHOR]

The article discusses the voting of the Swedish Parliament for the precedence of nuclear power development. The Swedish Parliament had a chance to go against the world trend toward nuclear power and lead the way toward a fission-free economy. The nuclear debate in Sweden began, as in many other countries, with the issue of nuclear weapons. Opinion against a Swedish nuclear weapons program was strong from the outset, and managed to keep nuclear weapons out of the Swedish arsenal. One consideration which led to the nuclear power program is that by investing in heavy water reactors (HWRs), the government hoped to avoid increased dependence on other countries for Sweden's energy supply.

The pulsed neutron source (PNS) technique was used to determine the prompt neutron decay constant for two different lattice pitches in the HWZPR heavy water zero power reactor. The results were compared to the variance-to-mean ratio (VTM) method. The neutron mean generation time was also calculated for both pitches, and the results were compared to previous Monte Carlo calculations. The findings of this research can be used as a benchmark nuclear codes to validate kinetic parameters. • Kinetic parameters of a zero-power reactor was studied in different lattice pitches by the pulsed neutron source method. • Prompt neutron decay constant was measured at different subcritical levels. • Critical prompt neutron decay constant extrapolated, modified, and validated with the variance to mean ratio (VTM) method. • Neutron mean generation time was obtained and compared with calculation results. • Effect of moderator to fuel ratio on the kinetic parameters was investigated. [ABSTRACT FROM AUTHOR]