Heavy Water Reactor Research Articles

Optimization of AHWR core design: neutronics and thermal-hydraulics analysis and calculations

Advanced Heavy Water Reactor(AHWR) is being designed by BARC for thorium utilisation and demonstrate all aspects of the thorium fuel cycle. The AHWR is a vertical pressure tube type thorium-based reactor cooled by boiling light water and moderated by heavy water. The equilibrium fuel cycle of AHWR is being optimised for a discharge burn-up of 40 GWd/Te with Plutonium as makeup fuel in a closed fuel cycle. The 233U content in the Uranium and fissile Plutonium content in the Plutonium have been assumed to be about 78% and 75% respectively. One of the important passive safety features of AHWR is heat removal through natural circulation which is governed by a strong neutronic and thermal hydraulic coupling. The neutronics design simulations were done for average coolant condition. In order to study the effect of natural circulation and its consequences on the coolant behaviour and the resulting power distribution, the neutronics and thermal–hydraulic coupled calculations were carried out. The neutronics calculations code (FEMFOL) and thermal–hydraulic calculations code (ARTHA) were externally coupled to estimate the coolant density distribution and the critical heat flux ratio (CHFR) for the fuel. This paper gives the details of the critical heat flux ratio estimation and the effect of density distribution on the core power distribution at some selected operating phases during core follow-up of AHWR.

Measurements on (233U, Th)O2 fuel by substitution experiments in the ZED-2 critical facility

This paper summarizes the results of experimental measurements on mixed-oxide (233U,Th)O2 fuel bundles in the ZED-2 heavy water critical facility of Canadian Nuclear Laboratories (CNL) in Chalk River, Canada. Pressure-tube heavy water reactors can use such fuel to extract the energy potential of thorium, which helps extend nuclear energy resources and reduces waste. Measurements were made in a series of substitution ZED-2 experiments, with (233U,Th)O2 test fuel placed into vertical channels filled with air or heavy water in a reference lattice of very low-enriched uranium fuel. Fuel was arranged to maximize the importance of the test fuel. Each fuel and coolant configuration in the substitution series was simulated in MCNP5 to verify the measurement of critical moderator height. The precision of the substitution measurements can help support the validation of reactor physics codes for evaluation of reactivity coefficients in reactors with fuel that contains oxides of thorium and thorium-derived uranium.

Safety Evaluation of Sustainable Uranium Development in China Combined with an Analytical GAN Framework

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.

Optimization of flux mapping in-core detector locations in AHWR using clustering approach

Measurements from in-core Self Powered Neutron Detectors in the nuclear reactor serve as input to the Online Flux Mapping System which is used for effective monitoring and regulation of the nuclear reactor. It is important to optimize the positioning of in-core detectors given that placement at all desired locations is not pragmatic. In Advanced Heavy Water Reactor (AHWR), which uses thorium-based fuel and boiling light water as coolant, the neutronics and thermal hydraulics are strongly coupled. However, neutronically different regions of the core may be loosely coupled. The knowledge of the 3-D core neutron flux distribution is an important safety feature of online monitoring in AHWR. This paper proposes a data-driven approach based on in-core detector measurements to arrive at the number and optimal locations for in-core detectors in AHWR. K-means clustering algorithm has been applied to in-core detector measurements of AHWR to identify the linear relationship among the detectors and to group them into clusters. Detectors within the clusters are highly correlated with each other and one detector is sufficient to represent others in a cluster, resulting in optimization of the number and locations of in-core detectors. The proposed in-core detector layout has been tested for its effectiveness in flux reconstruction using a Flux Mapping Algorithm in test cases representing typical operational scenarios of AHWR.

DOSE ASSESSMENT FOR ATMOSPHERIC DISCHARGE OF LONG-LIVED RADIONUCLIDES IN NUCLEAR POWER PLANT DECOMMISSIONING.

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.

The Current Status of the Heavy Water Detritiation Facility at the NRC (Kurchatov Institute) – PNPI

The multifunctional Tritium Removal Facility (TRF) has been designed for the heavy water research reactor PIK in Russia. Along with the extraction of tritium and protium from a heavy water reflector of the reactor, the TRF provides the processing of heavy water waste and the production of tritium-free heavy water. The combined electrolysis catalytic exchange process and hydrogen cryogenic distillation are used at the TRF. A number of investigations have been made to obtain data for the TRF design. At present, the facility is under construction and a detailed design is being completed. The possibility of processing heavy water waste to produce heavy water simultaneously with the extraction of tritium and protium from the heavy water reactor will improve the functionality and economic efficiency of the TRF. The basic technological scheme of the facility and the main modes of its operation are presented along with the progress of the construction of the facility building and engineering systems.

Calculation of Potential Radiation Doses Associated with Predisposal Management of Dismantled Steam Generators from Nuclear Power Plants

Although the generation of large components from nuclear power plants is expected to gradually increase in the future, comprehensive studies on the radiological risks of the predisposal management of large components have been rarely reported in open literature. With a view to generalizing the assessment framework for the radiological risks of the processing and transport of a representative large component—a steam generator—12 scenarios were modeled in this study based on past experiences and practices. In addition, the general pathway dose factors normalized to the unit activity concentration of radionuclides for processing and transportation were derived. Using the general pathway dose factors, as derived using the approach established in this study, a specific assessment was conducted for steam generators from a pressurized water reactor (PWR) or a pressurized heavy water reactor (PHWR) in Korea. In order to demonstrate the applicability of the developed approach, radiation doses reported from actual experiences and studies are compared to the calculated values in this study. The applicability of special arrangement transportation of steam generators assumed in this study is evaluated in accordance with international guidance. The generalized approach to assessing the radiation doses can be used to support optimizing the predisposal management of large components in terms of radiological risk.

Administrative dose control for occupationally-exposed workers in Korean nuclear power plants

Korean nuclear power plants (NPPs) have various radiation protection programs to attain radiation exposure as low as reasonably achievable (ALARA). In terms of ALARA, this paper provides a comprehensive overview of administrative dose control for occupationally-exposed workers in Korean NPPs. In addition to dose limits, administrative dose constraints are implemented to resolve an inequity of radiation exposure in which some individuals in NPPs receive relatively higher doses than others. Occupational dose constraints in Korean NPPs are presented in this paper with the background of how those values were determined. For pressurized water reactors, 80% and 90% of the annual average limit for an effective dose, 20 mSv/y, are set as the primary and secondary dose constraints, respectively. Pressurized heavy water reactors (PHWRs) have also established the primary and secondary dose constraints corresponding to 70% and 80% of the effective dose limit, and additional constraints for tritium concentration are provided to control internal exposure in PHWRs. Follow-up measures for exceeding these administrative dose constraints are also introduced compared to exceeding the dose limits. Finally, analysis results of dose distributions show how the implementation of administrative dose constraints impacted the occupational dose distributions in Korean NPPs during the years 2009–2018.

Special Section: Selected Papers From the Ninth International Symposium on Supercritical Water-Cooled Reactors

Special Section: Selected Papers From the Ninth International Symposium on Supercritical Water-Cooled Reactors

Kv-scaling analysis to support the validation of Atucha-II Best Estimate Evaluation Model

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 the Chapter 15 of FSAR (Final Safety Analysis Report) dealing with accident analysis. Approaches based on Best Estimate Plus Uncertainties (BEPU) to perform accident analysis of a Nuclear Power Plant (NPP) for licensing purposes request the availability of validated tools and specific procedures suitable for the analysis of accident conditions envisaged in the concerned NPP. This implies the necessity to adopt and to prove an adequate quality of the so-called Evaluation Model (EM), i.e. the ensemble of data, assumptions, codes and nodalizations developed and validated to be used for carrying out the safety analysis. The purpose of the present paper is to outline key aspects and results of one of the steps of the BEPU process concerning the qualification of the Atucha-II PHWR Evaluation Model for the licensing process. A special procedure, named “Kv-Scaled Analysis + Countepart Test Calculations” have been developed to take into account the specificities of Atucha-II NPP (CNA-2) and the fact that Atucha-II has not a prototype integral test facility. The proposed procedure uses a set of facilities which share some similar geometrical features and ranges of variations of relevant physical quantities for some subsystems or components of CNA-2, taking into account the different kind of accidents addressed in the Chapter 15 of the Final Safety Analysis Report (FSAR) of Atucha-II NPP. The final goal of the procedure is to qualify separately systems/sub-systems of the nodalization of the Atucha-II NPP while keeping the features of a full system analysis involving the whole NPP nodalization. A total of nine tests performed in six different scaled facilities have been considered and analyzed.

Reduced order model based FOPID controller design for power control in Pressurized Heavy Water Reactor with specific gain –phase margin

The substantial portion of the existing power generation comes from Pressurized Heavy Water Reactor (PHWR). However, PHWR is a nonlinear and high order system and designing a controller for such a high order system is not an easy task. Therefore, it is required to design a controller for a reduced order PHWR system which will give the same response as that on the original higher order system. The focus of the present research endeavor is to propose a design methodology for a fractional-order proportional integral derivative controller for the power control of reduced order PHWR system. The reduced order models of PHWR have been obtained using Balanced Truncation method and a robust FOPID controller has been designed using the stability boundary locus method with specific gain-phase margin. The simulation results illustrate a better set point tracking performance with the use of proposed FOPID controller at 30% control drop and varied initial power levels.

Integral Sliding Mode for Power Distribution Control of Advanced Heavy Water Reactor

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.

Provision of the radiation safety for the decomissioning of the heavy-water research nuclear reactor NRC «Kurchatov Institute» – ITEP

The article provides a brief description of organizational and technical measures aimed at ensuring radiation safety during the decommissioning of the heavy-water research nuclear reactor of Institute for Theoretical and Experimental Physics after A.I. Alikhanov of National Research Centre «Kurchatov Institute». Information is provided on the history and features of the operation of the reactor, including parameters and characteristics that are significant for planning and conducting work. The peculiarities of legal regulation in the field of ensuring radiation safety are given; regulatory acts and rules accompanying other activities during decommissioning and directly related to radiation safety are also considered. The paper describes the work done in preparation for dismantling, the initial and current state of the installation, forthcoming work with examples of dismantled equipment. Methods for handling radioactive waste arising during decommissioning are considered, including methods for fragmentation of large structural elements (examples of mechanical devices are given), methods for sorting according to different specific activity (high activity, low activity), radionuclide composition and physical properties (solid, metallic, non-metallic, liquid). A special method for handling liquid radioactive waste is described, which includes the collection and temporary storage system. To assess the radiation situation at workplaces during the dismantling of the reactor structures, calculations of radiation transfer were carried out on the running and shutdown reactor, during which it was established that the expected dose to the personnel when performing activities on decommissioning of TBR is much lower than the limit values, established by regulatory documents. In accordance with the estimated radiation doses, rules and instructions for personnel were determined, including the procedure for using personal protective equipment, the necessary measures for surface decontamination, etc. Information is given on the procedure for radiation monitoring at all stages of dismantling and at the final stages of decommissioning including control of premises, personnel, equipment, waste of various types, atmospheric air.

Development of PINN Controller for Fuel Handling System of Pressurized Heavy Water Reactors

Development of PINN Controller for Fuel Handling System of Pressurized Heavy Water Reactors

Dynamic reliability framework for a Nuclear Power Plant using dynamic flowgraph methodology

Passive safety systems are being considered in advanced reactor designs to provide inherent stability for the operation of the nuclear reactor. Passive shutdown is provided for guaranteed removal of decay heat under emergency conditions. The reliability of such systems should be very high. The static reliability assessment of such systems has been considered using various techniques such as fault tree analysis, failure mode effect analysis, and reliability block diagrams. Dynamic reliability methods are powerful mathematical frameworks capable of handling interactions among components and process variables explicitly. In principle, they constitute a more realistic modelling of systems for the purposes of reliability, risk and safety analysis. Although there is a growing recognition in the risk community of the potentially greater correctness of these methods, no serious effort has been undertaken to utilize them in industrial applications. The dynamic flowgraph methodology is an integrated methodological approach to modelling and analyzing the behavior of software-driven embedded systems for the purpose of reliability/safety assessment and verification. In the present work, dynamic flowgraph methodology has been used to analyze the Station Blackout Scenario for a Nuclear Power Plant. The benefits of the proposed method are brought out with respect to the traditional methods like fault tree analysis, which deteriorates in applicability with increasing system size and complexity and fails to accommodate the dynamics of the system. The proposed method has been validated on the passive residual heat removal system of pressurized heavy water reactor.

Experimental Evaluation of Critical Heat Flux in Downward-Facing Boiling on SS304 L Flat Plate Relevant to In-Calandria Retention in PHWRs

Abstract Retention of corium inside the calandria vessel (CV) by externally cooling it by calandria vault water is essential to mitigate severe accidents in pressurized heavy water reactor (PHWR). The thermal failure of CV can be prevented by effective decay heat removal on the outer surface of CV using vault water, which depends on the heat transfer behavior from the outer surface of CV to the vault water. Determination of limiting heat removal capability of vault subcooled water through outer surface of CV is very important. Since the CV has a very large diameter and length, the bottom most part of the CV almost behaves as a flat plate with downward natural convection boiling heat transfer. The natural convection heat transfer is lesser on the flat surface as compared to the curved surface of the CV. Thus, the critical heat flux (CHF) on the flat surface under downward boiling condition is the limiting CHF of the CV under external surface boiling scenario. In order to estimate CHF in this configuration with local boiling, experiments were carried out on a downward facing SS304 L flat plate simulating the conditions of CV of 700 MWel Indian PHWR. The pool boiling CHF obtained in this study is also compared with other earlier works.

Laser based technique for monitoring heavy water leaks in nuclear reactors: performance validation with conventional techniques

In a pressurized heavy water reactor (PHWR), monitoring leakages of heavy water from coolant and moderator systems is very important for reactor safety as well as health of operating personnel. In view of this, heavy water leak monitoring system based on off-axis integrated cavity output spectroscopy (OA-ICOS) technique is developed. The developed system estimates the amount of heavy water in the samples by measuring the HDO concentration with the detection sensitivity of ±1.2 ppm. The performance of this system is studied by measuring various samples obtained from the Dhruva research reactor at Bhabha Atomic Research Centre. The validation of the data obtained using this system is carried out using tritium activity measurements of same samples. The results are also compared with D/H ratio measurements obtained by isotope ratio mass spectrometer (IRMS) . The measured HDO concentration of various samples using the developed system shows a good correlation with the measurements using IRMS and tritium activity measurement techniques.

Changes of the abundances of Fe, Ni, Zr isotopes due to neutron radiative capture reactions in heavy water reactors

The abundances of the nuclides in nuclear materials change over time in reactors because of neutron-induced reactions like (n, γ), (n, α), (n, p), etc. Among them, the effect caused by (n, γ) reactions is especially significant in heavy water reactors because they have a large proportion of thermal neutron flux, and the cross sections of (n, γ) reactions for most nuclides are dominant in the thermal region than in regions of higher energies. The changes of the abundances of Fe, Ni, and Zr isotopes in CANDU-6 reactors over 25 years through radiative capture reactions (thermal neutron flux of 2.7 × 1018 n·m−2·s−1, and effects of epithermal, resonance, and fast neutrons are also included) are calculated. Results show that for Fe and Ni isotopes, the changes of abundances are obvious: 57Fe enriches from 2.12% to 12.22%; 58Ni drops from 68.09% to 36.88% and 60Ni increases from 26.23% to 44.35%. For Zr isotopes, the effect is also noticeable, with the abundance of 90Zr and 91Zr decreasing from 51.45% and 11.22% to 50.67% and 9.58%, respectively. . Present results are verified with the ALARA code. Based on these calculations, a new approach to estimating the neutron fluence in heavy water reactors is proposed according to the abundances of several stable isotopes including 56Fe, 57Fe, 58Ni and 60Ni.

Temporal evolution of activation products in a zircaloy–4 guide tube and the estimation of its external gamma dose rate for decommissioning activities

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.

Advanced heavy water reactor control with the aid of adaptive second-order sliding mode controller

PurposeAdvanced 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/approachThe 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.FindingsThrough 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/valueThis 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.