Advanced Heavy Water Reactor Research Articles

Impact of adding high-concentration neutron poisons to reactor moderator system for guaranteed shutdown

To ensure a reliable shutdown of the proposed advanced heavy water reactor (AHWR) of Indian origin during emergency conditions, it is imperative to use an increased concentration of liquid neutron poison for injection into the moderator system due to its higher reactivity. The sudden injection of this liquid poison may cause a surge in radiolytic product (H2O2 and H2) yield, posing safety risks. If H2/D2 yield exceeds 4 % (v/v), a combustible mixture with oxygen can be formed and presence of oxidizing H2O2 can enhance corrosion. Hence, accurate estimation of H2O2 and H2 is imperative from the reactor safety point of view. We investigated gadolinium and boron (two liquid neutron poisons) at varying concentrations through high energy electron beam and gamma radiolysis. The radiolytic yields from both the liquid neutron poisons of equivalent concentrations at a constant dose and operating moderator conditions in electron beam and gamma radiolysis were compared. The molecular product yield was observed to be concentration and dose-rate-dependent for both the neutron poisons. Electron radiolysis exhibited a higher yield of H2O2, whereas gamma radiolysis resulted in a greater H2 yield. In electron radiolysis, the H2O2 yield decreased with increasing B concentration. These results give an overall estimation of radiolytic products for a particular dose rate, total dose and effect of head space volume over liquid irradiation zone from these two liquid poisons of a particular concentration. These insights will aid in selecting suitable poisons and managing their effects effectively to avoid hydrogen deflagrations after emergency shutdown.

Development of a dual sensitive N,N,N’,N’,N”,N”-hexa-n-octylnitrilotriacetamide (HONTA) based potentiometric sensor for direct thorium(IV) estimation

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).

X-ray diffraction and Raman spectroscopic studies on ThO2−UO2 solid solutions

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, ThO2−233UO2 and ThO2−UO2 (low-enriched) mixed oxide fuels with < 30% UO2 have been proposed for the Advanced Heavy Water Reactor. In the present work, phase relations in the Th1-xUxO2+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 ThO2, 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.

Overview of the physics challenges in design of new energy systems

Today's energy systems have diverse objectives catering to enhanced safety, ease of operation and better fuel utilization both for power generation and several other applications of nuclear reactions. Bhabha Atomic Research Centre (BARC) has been engaged in the design of new energy systems such as Advanced Heavy Water Reactor (AHWR), Indian Pressurised Water Reactor (IPWR), Molten Salt Reactor (MSR) and High Temperature Reactors (HTRs). The design optimisation of these reactors has varied challenges as they involve new fuel, newer materials, complex geometries, and inherent safety. AHWR, MSR and HTRs based on thorium fuel cycle while IPWR is optimized for longer cycle length with enriched uranium. HTRs are being developed as nuclear power packs for remote areas and they have interesting physics features of long-life cores and hence more challenging to be modelled.New designs use new materials and therefore require updated thermophysical and neutronic properties. Neutron-nuclear interaction cross sections are required in uncharted regimes. Modelling of physical phenomena in an MSR or HTR is very different than in water-cooled reactors.The neutronic tools also are required to be made state-of-the-art in order to model these multi physics phenomena. An exercise was taken up for a coupled neutronic-thermal hydraulic simulations for the natural circulation based boiling water cooled AHWR. Several new algorithms were also developed for fuel management of advanced thermal reactors. Monte Carlo methods and improved deterministic methods have been continuously developed. This paper brings out the challenges in design of these advanced reactors and the need for new experiments and irradiations of new fuel in research reactors.

Benchmarking exercise of indigenous Monte Carlo neutron transport code PATMOC for hexagonal full core reactor geometry

An indigenous Monte Carlo neutron transport code, named PATMOC, has been designed and developed at Bhabha Atomic Research Centre (BARC) for reactor physics design applications. The code has been extensively tested for its algorithmic accuracy against numerous international numerical benchmarks, has also been validated on practical nuclear reactor systems like Advanced Heavy Water Reactor Critical Facility (AHWR-CF) and Apsara-U pool type research reactor located at BARC, and is now in regular use for day-to-day reactor physics studies. These reactor systems and most of the benchmark problems used for the verifications and validation of the code have square lattice arrangement in the core. In this paper, we present the results of verification of PATMOC code on whole-core reactor system with hexagonal lattice arrangement of fuel elements. In this exercise, we have analyzed the High Temperature engineering Test Reactor (HTTR) whole-core benchmark for integral and differential parameters at both lattice and core level. The results demonstrate the algorithmic accuracy of the code on hexagonal reactor systems also. This exercise will help make use of this code for reactor physics design studies also for the reactor systems with hexagonal arrangement.

Uncertainty and disturbance estimator based spatial power control of advanced heavy water reactor

Uncertainty and disturbance estimator based spatial power control of advanced heavy water reactor

Experimental study of voiding effects with thorium based MOX fuel cluster in Critical Facility

An Advanced Heavy Water Reactor (AHWR) based on thorium fuel is being designed and developed in India. The design has many inherent safety features; the most important among them is its negative coolant void coefficient. An experimental facility, named AHWR Critical Facility, has been built to validate the physics design parameters. Coolant void worth which is an important safety parameter was measured using differential critical height in a thoria based experimental MOX fuel assembly placed in the center of the core. The voiding effects were experimentally determined with both water and high density polyethylene (HDP) coolant. Different fractional voids were obtained using combinations of HDP block of compatible shape and sizes.The aim of the experiment was to validate the code systems used in physics design and in particular, to assess their capability of modeling the voiding conditions in thorium based fuel cluster. The validation exercise was performed with both deterministic and stochastic methods. The theoretical estimates of the void worth were found in very good agreement with that of the measurement. The results have enhanced the confidence in the code system used for AHWR physics design.

Design of fractional-order NPID controller for the NPK model of advanced nuclear reactor

Advanced heavy water reactor (AHWR) is a type of advanced nuclear reactor which is the most efficient nuclear fission reactor for power generation due to the large thorium reserve. Its safe, stable, and efficient operations are critical for the revival of the fission energy sector. Incorporation of a robust control scheme into the nuclear reactor model is required for proper trajectory tracking of demand power. The AHWRs are coupled, significantly nonlinear, and multi-input multi-output (MIMO) systems. External disruptions and time-varying characteristics harm the systems' performance. As a result, the controller built for these systems must be able to deal with the complexity, which is most challenging for control engineers. A fractional-order nonlinear proportional, integral, and derivative (FONPID) control method is presented in this study for normalized power distribution management of the AHWR using normalized point kinetic equations (NPKEs) for trajectory tracking, disturbance rejection, and noise suppression tasks to improve the output power. All controller settings are fine-tuned using a genetic algorithm (GA) with the sum integral of time and square error (ITSE). The suggested FONPID controllers' performance is compared to its integer-order control structure, i.e., NPID and classical PID control structure. External disturbances at controller output and random noise at the sensor output are tested for resilience to establish the usefulness of the suggested control methods. The simulation results showcased that the proposed FONPID controller outperforms its integer-order (IO) counterpart as well as the traditional PID controller.

Studies on the behaviour of a passive containment cooling system for the Indian Advanced Heavy Water Reactor

Abstract A passive containment cooling system has been proposed for the advanced heavy water reactor being designed in India. This is to provide long term cooling for the reactor containment following a loss of coolant accident. The system removes energy released into the containment through immersed condensers kept in a pool of water. An important aspect of immersed condenser’s working is the potential degradation of immersed condenser’s performance due to the presence of noncondensable gases. An experimental programme to investigate the passive containment cooling system behaviour and performance has been undertaken in a phased manner. In the first phase, system response tests were conducted on a small scale model to under­stand the phenomena involved. Tests were conducted with constant energy input rate and with varying energy input rate simulating decay heat. With constant energy input rate, pressures in volume V1 and V2 reached almost steady value. With varying energy input rate V1 pressure dropped below the pressure in V2. The system could efficiently purge air from V1 to V2. The paper deals with the details of the tests conducted and the results obtained.

Experimental studies on condensation of steam mixed with noncondensable gas inside the vertical tube in a pool filled with subcooled water

Abstract A passive containment cooling system with immersed condensers has been proposed as one of the alternatives for the Advanced Heavy Water Reactor (AHWR) being designed in India. The system removes residual/decay heat released into the containment through the immersed condensers kept in a pool of water following loss of coolant accident. An important aspect of the immersed condensers is the potential degradation of its performance due to the presence of noncondensable gases. Experiments are performed to obtain reliable data on condensation phenomena in presence of air. These experiments are conducted on full-scale tubes of condensers immersed in a pool of water maintaining similar conditions as in the prototype of AHWR. A method has been proposed for the determination of the local heat transfer rate using correlations given in literature. The parametric effects of air mass fraction, pressure, steam flow, etc. on condensation heat transfer in presence of noncondensable gas have been studied. The experimental results are compared with the correlations given in literature.

Sensitivity analysis of AHWR flux mapping system for unavailability of SPNDs

In Advanced Heavy Water Reactor, a set of 168 Self-Powered Neutron Detectors (SPNDs) distributed uniformly at 6 axial locations in 28 vertical In-core Detector Housing (ICDH) units each, has been proposed to be used for monitoring and control of neutron flux distribution in the reactor core. These 168 SPNDs arrived at by applying the K-means clustering approach on the simulated in-core detector measurements, proved to be effective in flux reconstruction during normal reactor operational scenarios such as the spatial flux variations caused by movement of regulating rods, xenon spatial oscillations, and on-power refuelling. However, the accuracy of flux mapping may be affected in case of the unavailability of one or multiple SPNDs. Hence the effect of the unavailability of SPNDs on accuracy characteristics of flux mapping needs to be studied. Motivated by this, the performance of the flux mapping system in case of unavailability of all 6 SPNDs of an ICDH unit under various nominal reactor operational scenarios has been studied. It is evident from the analysis results presented in the paper that though the RMS error is poorer compared to the case when all SPNDs are available, the overall RMS error is better than 7%. The studies also brought out that the ICDH units located in the outer region of the core have more bearing in the flux reconstruction.

Performance evaluation of AHWR flux mapping system during normal operational scenarios

In large nuclear reactors, a number of appropriately distributed in-core neutron detectors are used for monitoring and control of neutron flux distribution in the core. In Advanced Heavy Water Reactor (AHWR) too, a set of 168 self-powered neutron detectors installed vertically at 6 locations in 28 in-core detector housing units has been proposed to be used. This configuration arrived at using the K-means clustering approach, has been found to reconstruct the core neutron flux distribution for a number of steady-state flux shapes. For practical application, in addition to steady-state, the optimized set of in-core detectors should provide reasonably accurate core monitoring for regular reactor operations as well. This paper investigates the accuracy characteristics of flux mapping using the proposed set of 168 in-core self-powered neutron detectors in the reconstruction of time-varying flux shapes encountered during normal reactor operations. Time-dependent flux shapes arising due to perturbations such as the movement of reactivity devices, refuelling and xenon spatial oscillations are considered for this study. Flux reconstruction is performed using the well-known flux synthesis method. From the analysis, it can be observed that the proposed set of in-core detectors reliably reconstructs the flux distribution in the reactor even during the normal operational scenarios. The maximum RMS error in mesh flux using 168 SPNDs and 35 eigenmodes is found to be 4.45% across all the operational scenarios considered.

Quantification of boiling flows in single and multiple heater rods assembly by recurrence plots and recurrence quantification analysis

Tracer particle position time series obtained by RPT experiments were used to quantify the changes in hydrodynamics induced by boiling flows in single and multiple rods assembly. The test section under consideration is a “scaled-down” version of a typical boiling water nuclear reactor (Advanced Heavy Water Reactor (AHWR) of the Department of Atomic Energy, India). The analysis of the above-mentioned time series was done by Recurrence Plots (RP) and Recurrence Quantification Analysis (RQA). Visual differences in the RPs were detected with changes in the inlet condition and heater rods configuration. The constructing feature of the recurrence plot can be categorized in two forms: local white areas (LWA) and local bold areas (LBA). However, the change cannot be quantified by visual observation, and for this quantification RQA parameters such as Recurrence Rate (RR), Determinism (DET), and Entropy (ENT) were used. Experimental results show that the increase in liquid flow rate, decrease in inlet liquid temperature, and increase in the number of the heater rods caused increased RR, increased DET, and increased ENT.

Heat transfer characterization under radial jet and falling film induced rewetting

Abstract Empirical correlations are developed for rewetting velocity and maximum heat transfer coefficient during rewetting phase of single hot vertical Fuel Pin Simulator (FPS) by using radial jet impingement and falling film. Emergency Core Cooling System (ECCS) has been designed for Advance Heavy water Reactor (AHWR) to rewet the hot fuel pin under the loss of coolant accident. Coolant injection takes place from a water rod which is located at the center of the fuel bundle in form of jets to rewet hot surface of fuel pin under loss of coolant accident. This kind of design to reflood the fuel bundle is different than bottom and top spray reflooding practiced in PWR and BWR type of nuclear reactors. There are two different kinds of rewetting found during radial jet induced cooling. The first one is due to radial jet impingement and the second one is due to falling film which is below the jet impingement point. Rewetting velocity has been predicted along the length of fuel pin due to radial jet impingement cooling. Temperature of FPS has been varied from 400°C to 700°C with help of different powers supply, simulating decay heat of reactor. A variation of coolant radial jet mass flow rate is from 0.5 lpm to 1.8 lpm. It is considered during ECCS injection. It has been observed from the experiments that rewetting velocity decreases with increasing the clad surface temperature and increases with increasing the coolant mass flow rate. The rewetting velocity in falling film is found to be nearly 1.8 times higher than rewetting velocity predicted in circumferential direction. Further, it is found that maximum heat transfer coefficient increases with increasing the radial jet coolant mass flow rate. The maximum heat transfer coefficient in case of radial jet impingement is found to be nearly 1.5 times the falling film rewetting. Developed correlation predicts the maximum heat transfer coefficient with experimental data well within the error band of ±10%.

Development of computer code ADWITA and data library for the solution of transmutation chain equations and application to the analysis of nuclear fuel cycles

A fuel cycle analysis code, ADWITA (Activation, Decay, Waste Incineration and Transmutation Analysis) has been developed which can generate isotopic inventory, calculate radioactivity, and decay heat based on irradiation history. The code can be used for fuel cycle design and transmutation studies. The system of linearized transmutation chain equations is solved by application of Bateman series solution method and the matrix exponential method in which the matrix exponential is approximated by the exponential power series. The required data libraries for Pressurised Heavy Water Reactor (PHWR) and Advanced Heavy Water Reactor (AHWR) have also been developed. The code and the data have been validated using the analytical decay of 237Np, the PHWR fuel cycle benchmark data available in IAEA TECDOC and experimental measurement of isotopic composition and decay heat in AECL report. The code has been applied for the analysis of fuel cycle of AHWR which has been designed for the utilisation of thorium as nuclear fuel on large scale and envisages fuel reprocessing. The application of code ADWITA and its library to AHWR has helped to understand the unique features of AHWR fuel cycle. It has been found that in AHWR the activity and decay heat due to actinides and fission product (FP) in both (Th, 233U) MOX and (Th, Pu) MOX fuel is similar at the discharge but the activity and decay heat due to actinide takes over that due to FP in case of (Th, Pu) MOX earlier. The accumulation of uranium isotopes at the end of burnup in individual (Th, 233U) and (Th, Pu) MOX fuel pin has been estimated which is an important input to the reprocessing of these fuels. The rise in activity after reprocessing due to presence of 232U has also been estimated.

A multiscale data reconciliation approach for sensor fault detection

When used separately on the sensor data of the processes like nuclear reactors, the data reconciliation with fault detection and isolation strategy gives noise-corrupted estimates, and the wavelet transformation gives erroneous inferences about the operating point of the process under sensor fault conditions. Aiming to solve these challenging problems, a hybrid multi-scale data reconciliation scheme that combines data reconciliation with the wavelet transform is proposed in this work. The proposed method uses the steady-state data reconciliation framework under the assumption of consistent algebraic relationships among the wavelet coefficient data. The role of multivariate techniques in obtaining the algebraic relationships, online detection and isolation of sensor faults, orthogonal decomposition, and reconciliation of the wavelet coefficients data is demonstrated. It is shown that the reconciled estimates obtained from this method very closely represent the true behavior of the process as problems with respect to random noise, high-frequency components due to process faults, sensor faults, and the influence of sensor faults on the signal estimates are alleviated. The effectiveness of this method is quantitatively established when applied to the ex-core neutron detector data of the advanced heavy water reactor in various simulations.

Advanced Heavy Water Reactor: A New Step Toward Sustainability

One of the great advances in the current evolution of nuclear power reactors is occurring in India, with the Advanced Heavy Water Reactor (AHWR). It is a reactor that uses thorium as part of its fuel, which in its two fueling cycle options, in conjunction with plutonium or low enriched uranium, produces energy at the commercial level, generating less actinides of long half-life and inert thorium oxide, which leads to an optimization in the proportion of energy produced versus the production of burnt fuels of the order of up to 50%. The objective of this work is to present the most recent research and projects in progress in India, and how the expected results should be in compliance with the current sustainability models and programs, especially the "Green Chemistry", a program developed since the 1990s in the United States and England, which defines sustainable choices in its twelve principles and that can also be mostly related to the nuclear field. Nevertheless, in Brazil, for more than 40 years there has been the discontinuation of research for a thorium-fueled reactor, and so far there has been no prospect of future projects. The AHWR is an important example as an alternative way of producing energy in Brazil, as the country has the second largest reserve of thorium on the planet.

Evaluation of elevated temperature process for the decontamination of stainless steel 304LN surface

Abstract Deposition of activated corrosion products on out-of-core surfaces and ensuing radiation field build-up in nuclear reactors calls for remedial measures like decontamination to minimize exposure of service personnel carrying out routine maintenance jobs. Decontamination of stainless steel structural materials in nuclear power plants pose challenges due to the inherent difficulty in dissolving chromium containing oxides like ferrites or chromites. Stainless steel, SS 304LN, is the proposed structural material for the main heat transport system of Indian Advanced Heavy Water Reactors (AHWRs) and it is important to be equipped with an effective decontamination process to mitigate the radiation field build-up in AHWRs. This paper discusses the studies to develop an elevated temperature (ET) decontamination process for SS304LN surface. Oxide covered SS304LN specimens were prepared under simulated AHWR coolant chemistry conditions. The oxide film formed on the specimen surface was Ni and Cr substituted single phase ferrite spinel. The oxide film could be completely dissolved from the SS304LN surface in the nitrilotriacetic acid (NTA) based ET process at 180 °C in a single step with an acceptable and low corrosion rate of 7.7 × 10−3 μm/h. The efficiency of the ET process was compared with the known NTA based conventional low temperature dilute chemical decontamination process and a multi-step, multi-cycle process comprising of alternate oxidation and reduction processes. The ET process proved to be more effective and apt for decontamination of SS304LN surface.

Development of in-core fuel management tool for AHWR using artificial neural networks

In-core fuel management methods based on evolutionary algorithms and machine learning tools are being developed worldwide to improve the overall efficiency of the operating cycle in all type of nuclear reactors. The fuel cycle of Advanced Heavy Water Reactor (AHWR) is unique and requires complex fuel management involving in-core refueling and reshuffling operations. Using the Artificial Neural Networks (ANN) approach, we have evolved an effective fueling strategy for AHWR. A computer code based on ANNs has been developed and predictions are done for k-effective and maximum channel power for all possible refueling inputs. The best candidates are chosen to perform 3D diffusion simulations. Unlike LWRs, the AHWR has on-power refueling feature which results in continuously changing core configuration and core burn-up profile. The paper highlights the use of ANNs for arriving at an optimized refueling strategy for AHWR and makes a comparison with earlier results where heuristic approach was used.

Nonhyperbolicity of Conservation Equations of RELAP5 Two-Fluid Model in Nuclear Reactor Safety Results (Investigation and Eigenvalue Analysis)

Abstract The RELAP5 code simulates the thermal-hydraulic characteristics of nuclear reactors by the use of a two-fluid one-dimensional, nonequilibrium, nonhomogeneous two-phase flow model. This model consists of six governing equations to describe the mass, energy, and momentum of the two fluids. The scope of this work comprises the study of the mathematical nature of the code model and to predict the accuracy of the model in the nuclear reactor safety analysis. The method of characteristics (MOC) is applied to check the nonhyperbolic nature of conservation equations for all normal and accident conditions of light water reactors (LWRs). The analysis also gives information about the soundness of the model and to identify the regions where the solutions obtained from it will be numerically convergent. The characteristics of equations of nonhyperbolic nature are complex. It implies that results thus obtained (by finite difference method) have to be interpreted very carefully in view of the sensitive nature of reactor safety analysis. The present analysis shows that governing equations of the code exhibit complex characteristics for some operating conditions thus implying nonhyperbolicity under those conditions. Results are less accurate under such conditions, so sensitivity analysis plays an important role. The sensitivity of closure relationship on the conservation equation's stability is also checked. The analysis is performed in matlab environment for three different systems, (a) pressurized water reactor (PWR), boiling water reactor (BWR), and (c) natural circulation reactor or advance heavy water reactor (AHWR). These results can also be extended to other thermal-hydraulic systems. The different values of the coefficient of closure relationship are taken for different flow regimes. It is observed that the coefficient of virtual mass (for momentum equation) has a significant effect on the hyperbolicity of the system. It is recommended that further development of the RELAP5 model be performed to identify changes that would reduce the region of complex characteristics. The importance of MOC (in nuclear reactor thermal-hydraulic safety analysis) is evident here. In addition, a detailed analysis for operating pressures range of 0.1–22.5 MPa is also performed to find out the nonhyperbolic regions of code model and realistic data of the different type of reactors is used as input of the code. It is also observed here that RELAP5 results are less accurate when system pressure exceeds 19.5 MPa.