Heavy Water Reactor Research Articles

Neutronic analysis of high burnup thorium-HALEU fuels in PHWR

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.

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).

Field tests for performance evaluation of a new spent-fuel verification system in heavy water reactor

The optical fiber-based measurement system has been utilized for inspecting CANDU-type spent fuel at the Wolsung site in South Korea under the nuclear safeguards of the International Atomic Energy Agency (IAEA). In our previous studies, we developed a new spent-fuel verification system to address issues of the existing instrument. While our prior work primarily focused on equipment development, we had not yet conducted comprehensive field tests to assess its performance. Recently, we conducted field tests at the Wolsung Unit 4 to thoroughly evaluate the newly developed spent-fuel verification system's performance. This paper discusses the results of these field tests, with a focus on signal sensitivity, the system's capability to distinguish neighboring spent-fuel bundles, and the optical fiber background signals. Additionally, we conducted experiments to assess how different scintillation materials impact the system's performance, including p-terphenyl organic scintillator, PVT plastic scintillator, and lithium glass scintillator. The experimental results demonstrated that the new instrument equipped with the p-terphenyl organic scintillator outperformed the existing system. The p-terphenyl scintillator emerged as the superior choice among the three radiation scintillators. Signals that were previously undetectable using the Li glass scintillator were observed in the signals obtained with the p-terphenyl and PVT plastic scintillators. The remarkable performance of the new verification instrument is attributed mainly to the p-terphenyl's high light output and low decay time. The newly developed verification system is expected to streamline IAEA safeguards inspection efforts, reducing both time and the burden on nuclear operators. This report represents the first successful application of the p-terphenyl scintillator as a radiation detector in a wet spent fuel pool. The experimental findings presented in this paper are anticipated to be valuable for researchers working on radiation detectors suitable for high radiation environments in water.

Study of grain growth kinetics of Zr-2.5%Nb alloy quenched from (α+β) and β-phase region

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.

Moisture Determination in CO2 Gas Cylinders for Application in Annulus Gas Monitoring Systems of Pressurized Heavy Water Reactors (PHWRs)

Background:: Critical monitoring of moisture content in annulus gas is required due to its detrimental effect on the pressure tube, leading to surface corrosion and finally the rupture of the pressure tube. Methods and Results:: The present work describes a simple and reliable thermo-analytical method for the quantitative determination of moisture in CO2 gas cylinders, using anhydrous barium chloride as the novel adsorbent material and simultaneous. Thermogravimetry - Differential Thermal Analysis - Evolved Gas Analysis (Fourier Transformed InfraRed Spectroscopy) [TG-DTA-EGA(FTIR)] as determination techniques. Conclusion:: Conclusion: Differential Scanning Calorimetry (DSC) coupled with Microscopy has been used to see the surface morphological variations during the heat treatment. The results of these studies are being discussed here.

Operating Experience of Snubbers in Indian Pressurized Heavy Water Reactors

In a Nuclear Power Plant (NPP), the piping layout needs to have flexibility to reduce the operating stresses and nozzle loads due to thermal expansion. At the same time, the piping and equipment need to be supported adequately to sustain the dynamic loads viz., seismic loads, sudden opening/closing of valves etc. In Indian Pressurized Heavy Water Reactors (PHWRs), Snubbers are provided on nuclear equipment & piping to serve both purposes i.e. allow flexibility for thermal movement during operation and at the same provide requisite rigidity to take care of dynamic loads. Snubbers are subjected to harsh environment during operation including high temperature, radiation, vibration and corrosive condition, which may adversely affect its performance. Various degradations viz., high drag force, shift in dynamic characteristics (locking characteristic), jamming etc. have been observed. Periodic testing of Snubbers is required to ensure the proper functioning during normal operating condition as well as during dynamic event. A dedicated In-Service Inspection (ISI) program for snubbers exists in Indian PHWRs to ensure healthiness/functionality of Snubbers. Efforts have been made to reduce numbers of Snubbers progressively in PHWR units. Numbers of snubbers have been significantly reduced in latest 700MWe PHWR Design. This paper covers the various degradations observed in snubbers, ISI program for snubbers, design modifications, Snubber testing, assessment for fitness for service, replacement of Snubbers in older units and optimization of snubbers.

Fretting wear assessment in PHWR pressure tubes using transient dynamic FEA and Archard’s wear model

In Pressurized Heavy Water Reactors (PHWRs), the fuel bundles are placed in a pressure tube (PT). These bundles rest on inside of pressure tubes through the bearing pads. The pressure tubes are part of the primary pressure boundary of the reactor, and are crucial from functionality and safety point of view. Any phenomena/ factors that may cause structural degradation of PT and may affect its structural integrity, requires thorough investigation. The fretting of pressure tube at bearing pad locations is one such phenomenon, wherein the local material removal due to fretting may lead to stress concentration and associated structural degradations. Therefore, a detailed assessment of the PT fretting damage incurred during its intended life is required. Present paper covers the premise and methodology based on Archard’s wear model for evaluation of the volumetric wear rate of pressure tube. A case study has been done to evaluate the fretting wear in pressure tubes of a large Indian PHWR. 3D transient computational fluid dynamic analysis and finite element analysis with contact modelled between the fuel bundle bearing pads and pressure tube are carried out to evaluate the contact force and sliding distance time histories. Based on the analyses results, the work rate and average saturated fret depth are evaluated. The assessment revealed insignificant concern owing to the fretting of pressure tube for the considered cases.

An analysis method to assess in-Calandria corium retention time in PHWRs

In Pressurized Heavy Water Reactors (PHWRs), during a postulated severe accident involving the core collapse, the core internals would fall onto the Calandria bottom and form a debris bed. Under such conditions, the core debris bed would be contained by the Calandria, which also facilitates its cooling by the Calandria vault water. In absence of any accident management action, the Calandria temperatures would rise, causing loss of material strength that may also threaten the structural integrity of Calandria. The gross structural failure of Calandria would limit its ability to further retain the core debris and would jeopardize the in-vessel accident management strategies. The knowledge of structural degradation of Calandria with time till its ultimate failure is important for formulating and assessing the severe accident management guidelines. Assessment of Calandria failure under accident conditions involves several important aspects such as complex thermo-mechanical loadings, temperature dependent elastic-plastic and creep modeling, flexibility of different parts and boundary constraints etc. In view of this, a systematic analysis method based on the sequentially coupled thermo- mechanical analysis has been devised and used for the structural assessment of Calandria. The failure assessment has been carried out as per the specified failure criteria for prominent failure modes including large unbounded deformations and plastic instability, large inelastic strains/ ductility exhaustion and creep-stress rupture. The paper covers various considerations taken into account during the analyses along with few case studies on Indian PHWRs.

Effect of Hydride Morphology on Fracture behavior of Zr-2.5%Nb Pressure Tube Material

Hydrogen generated due to corrosion reaction is partly absorbed by Zr-2.5%Nb pressure tube (PT) material during service in the Pressurized Heavy Water Reactors (PHWRs). Hydrogen present in excess of the solubility limit get precipitated and embrittles the host PT material. The degree of embrittlement of the host PT matrix is governed by the morphology of the hydride precipitates. An experimental setup was designed, fabricated and used to form hydrides of different morphologies using internal pressurization. The internal pressure during the treatment was sufficient to induce hoop stresses more than the threshold for stress for re-orientation of hydrides in the PT spool. The design of the setup ensured constant pressure inside the PT spool while cooling from hydrogen solutionizing temperature. Metallographic examination of the hydrogen charged PT spool after the reorientation treatment revealed formation of stress reoriented hydrides. Series of burst tests were carried out on the PT spools with varying hydride morphology. The burst pressure as well as the final crack extension were found to be function of hydride morphology. This paper summarizes multiple full scale burst tests performed on as received Zr-2.5%Nb PT spools as well as on PT spools with varying hydride content and morphology.

Kinetics of hydrogen absorption/desorption in the Zr-2.5Nb alloy.

The Zr-2.5Nb alloy is a typical pressure tube material in heavy water nuclear reactors, and an increase of hydrogen isotope content in the alloy during service can pose major safety risks; hot vacuum extraction-mass spectrometry is an efficient method for evaluating hydrogen isotope concentrations in the Zr-2.5Nb alloy. This work investigates the kinetics and thermodynamic properties of deuterium (D) absorption and desorption of the Zr-2.5Nb alloy using the constant volume adsorption method and the hot vacuum extraction method. In addition to the previously reported volume contraction model, it was observed that at 600 °C and above, the reaction between D2 and Zr-2.5Nb is dominated by diffusion, while the reaction is predominantly influenced by surface adsorption and dissociation below 600 °C. Phase transition sequence of Zr-2.5Nb deuterides during non-isothermal desorption was established using quantitatively calibrated thermal desorption spectra combined with the phase transition process of deuteride decomposition. These results can provide important references for optimizing the process parameters of the hot vacuum extraction-mass spectrometry method.

Structural Integrity Assessment and Fatigue Evaluation of Feed Water Nozzle of 700MWe PHWR Steam Generator

Steam Generator (SG) is a prime component in Indian Pressurized Heavy Water Reactors (PHWRs). The Steam Generator consists of an inverted vertical U-tube bundle in shell for primary flow (Heavy water). The shell side is designed for recirculation. Steam Generator is a Safety Class-1 component and designed & fabricated as per the requirements of ASME Section III, Subsection-NB.Secondary side of the steam generator consists of a shell, Feed Water Nozzle (FWN), Steam Outlet Nozzle, moisture separator, drier, tubes and shroud. The principal material of construction for secondary side pressure boundary is of low alloy carbon steel. Secondary side of steam generator is subjected various loadings (Dead weight, pressure, pipe reactions & thermal) during normal plant condition as well as transient conditions including severe transient like crash cool down where in the thermal gradients are much higher. To take care of these thermal transient loadings a special thermal sleeve type of design has been incorporated in the FWN.As mentioned above, FWN is subjected to thermal transients throughout its life and susceptible to high thermal gradient which makes it important to evaluate for structural integrity and fatigue failure. This paper covers the approach & results for structural integrity assessment (stress evaluation) and fatigue evaluation (cumulative fatigue damage factor) of Feed Water Nozzle of 700MWe PHWR SG for various Service levels A, B, D & Design. These service conditions comprise of thermal & pressure transients during operating and accidental conditions. Detailed integrity check performed for all conditions by stress categorization at critical locations and qualify primary & secondary stresses against ASME Section III allowable.

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.

CFD analysis of PHWR exposed core under postulated severe accident condition

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.

Reconstruction error‐based fault detection of time series process data using generative adversarial auto‐encoders

AbstractFaults in time series process data are typically difficult to detect due to the complex temporal correlations of data samples. In this context, traditional unsupervised machine learning algorithms, such as principal component analysis, independent component analysis, and so forth, would yield only limited performance. Deep learning‐based methods have been employed in recent years to address these problems. Recently, generative adversarial networks have emerged as a promising generative modelling approach for learning data distributions. Inspired by the above, in this study, we present a novel reconstruction error‐based fault detection method for time series process data using generative adversarial auto‐encoder (GAAE) with Wasserstein loss, cycle consistency loss, and gradient penalty methods. The proposed method is designed to detect abnormal patterns in the time series data by training a GAAE to learn the underlying normal behaviour of the data. GAAEs help in effectively capturing the data's hidden distribution, and Wasserstein loss with gradient penalty is used to improve the accuracy of the latent space representation of the data, while the cycle consistency loss ensures consistency between the input and output data during the reconstruction process. Based on the extent of the reconstruction error metric of the GAAEs, we identify the potential faults in the process data stream. The proposed method is evaluated on two data process sets, namely, the Tennessee Eastman benchmark process dataset and the nuclear power flux real‐time dataset from a pressurized heavy water nuclear reactor, to validate the efficacy of the proposed approach.

Analytical model development for investigation on hydrogen management studies in the containment of a nuclear reactor

Containment is the ultimate physical barrier to prevent the spread of active air-borne fission products as well as radiation shielding in normal operation as well as accident condition and is designed to enclose whole reactor systems. The design of containment system for Fleet mode of 700 MWe Indian Pressurized Heavy Water Reactors (IPHWRs) incorporates a Primary Containment (PC) of pre-stressed concrete enveloped by a Secondary Containment (SC) of reinforced concrete. For the specification of the design pressure of the containment system, analysis is performed for various postulated accidents such as Loss of Coolant Accident (LOCA) and Main Steam Line Break Accident (MLSB). Further Studies on hydrogen generation and distribution for various accident scenarios in reactor containment of Indian Pressurized Heavy Water Reactors (IPHWR) indicate that during severe accident with core damage, the possibility of global deflagration in absence of mitigating features is high and may challenge the integrity of containment. Therefore, it is essential to incorporate mitigating provisions inside the containment to prevent the hydrogen reaching local detonation and global deflagration during such a scenario.In order to arrive at the design basis of containment system and simulation of large numbers of accident sequence a computer code is required, where mitigating features such as Passive Auto-Catalytic Recombiners, Containment Spray System & Containment Filtered Venting System can be modeled. A system thermal hydraulic computer code has been developed for containment response calculation during normal operation as well as accident conditions including, severe accident to include the new models for mitigating features. This article presents the details of various models of computer code PACSR-SI-2.0, its validation with data observed for hydrogen distribution & removal in Recombiner Test Facility,Battelle Spray Test for containment system & Spray test at IIT Mumbai.

Effective methods for monitoring internal contamination among workers handling radioactive materials in various fields

This study aimed to develop effective methods for monitoring internal contamination among workers handling radioactive materials in various fields. A total of 160 workers from nuclear power plants, medical institutions, military units, and educational/research institutions were included in the study. The monitoring methods included urinalysis and whole-body counting (WBC) using a mobile radio bioassay laboratory (MRL). Gamma-emitting radionuclides were monitored using the MRL WBC system, and a separate pretreatment procedure was used for tritium measurement in urine samples. Gross beta-screening was performed using a liquid scintillation counting system. The results were evaluated on the basis of the established screening criteria and compared with the dose limits. Additionally, tritium concentrations in the bodies of workers in the vicinity of a heavy-water reactor was analysed to assess the association between tritium concentration and occupation. The results showed a wide distribution of tritium concentrations. Workers involved in fuel and maintenance tasks demonstrated the maximum exposure. Workers in medical facilities showed low levels of internal contamination, which was primarily related to tasks involving radioactive isotopes. Military personnel involved in equipment repair showed significant tritium contamination due to damage during repairs. Workers in educational and research institutions in general had low levels of internal contamination.

Computational Fluid Dynamics Calculations of Moderator Heat and Fluid Flow of Small Modular Heavy Water Reactor

Abstract The heavy water reactor concept Teplator is a pressure channel type reactor with independent systems for the primary coolant and the moderator. The present study analyses the low-pressure moderator cooling system of Teplator during full-power operation. The moderator is heated from neutron thermalization, gamma rays absorption, fission product decay, and decay of activation products. Additionally, heat transfer from the coolant channels has to be taken in the analyses of the moderator cooling system. Preliminary thermal-hydraulic analyses of the cooling system are supplemented by computational fluid dynamics (CFD) simulations of heat and fluid flow in the moderator's vessel, emphasizing flow-type regimes. Results from CFD simulations showed that the buoyancy-dominated flow (case MF-22) resulted in a higher thermal stratification and high moderator temperature close to the upper plate of the moderator vessel. The inertia-dominated flow regime MF-90 resulted in good mixing of the moderator and a low thermal stratification in the vessel. Finally, the midmass flow rate regime MF-45 was identified as a transitional region from a buoyancy-dominated to a mix-type regime.

Effect of temperature on the delayed hydride cracking rate of Zr-2.5Nb alloy pressure tubes

Delayed hydride cracking (DHC) is a potential failure mechanism of Zr-2.5Nb alloy pressure tubes, employed in heavy-water reactors. This study focused on understanding the DHC behavior of the Zr-2.5Nb pressure tubes within the temperature range 393–523 K. Our findings demonstrate that within this temperature range, the delayed hydride cracking rate (DHCR) and temperature follow the Arrhenius relationship, with the activation energy of crack propagation is found to be 43.16 kJ/mol. The changes in the fracture morphologies at different temperatures are analyzed. Variations in the fracture hydride lengths with temperature and the stress intensity factor K are obtained. Furthermore, it is found that the influencing mechanism of temperature on DHC behavior changes as the temperature increases.

Case studies on radioactivity and decay heat contribution from coating materials of irradiated accident tolerant fuel for a typical PHWR

Seven coating materials such as Al2O3, NbC, Si, SiC, TiN, Y2O3, and ZrC have been considered as a possible coating material towards development of accident tolerant fuel for a typical pressurized heavy water reactor. For an effective severe accident management strategy in addition to spent fuel handling, storage and transport, it is required to know the contribution to the radioactivity burden and decay heat load from the irradiated coating materials of accident tolerant fuel. Under irradiation of coating material in PHWR flux condition, the decay chain of each activated radionuclide has been simulated using ISOGEN code using one group homogenized point cross section. In this paper, radioactivity and decay heat contributions from different irradiated coating materials have been quantified and compared. The minimum contribution to radioactivity and decay heat is observed in SiC after irradiation of 445 days of normal operation. Out of all the coating materials considered in the study, most radioactivity and decay heat were added due to irradiation of coating materials Y2O3. From reactivity load, source term, and decay heat perspectives, silicon-based coating materials are found to be the most suitable.

Influence of chromium of low content on the corrosion behavior of SA106B carbon steels in high-temperature pressurized LiOH medium

The corrosion of carbon steel (CS) pipelines widely troubles the primary circuit system of pressurized heavy water reactors. The element chromium (Cr) plays a crucial role in resisting corrosion. However, the mechanism of oxide film formation and dissolution on low Cr-content CS remains unclear. A comparison study in corrosion behavior between SA106B with 0.02 wt% and 0.36 wt% Cr was carried out with weight loss tests, electrochemical measurements, and morphology observations. Immersion tests were performed in deaerated lithium hydroxide solution at 553 K and 6.3 MPa for 30 days. The results showed that although the content of Cr was as low as 0.36 wt%, it exhibited relatively high corrosion resistance to the SA106B with 0.02 wt% Cr. Two layers of oxide film were formed on both the two CSs and chromium oxide was present in the inner oxide layer of 0.36Cr CS. The possible formation mechanism of oxidation film with low content of Cr in high-temperature water was discussed.