Heavy Water Research Reactor Research Articles

Radiation source terms analysis and classification of radioactive waste for the decommissioning of the first HWRR reactor in China.

Nuclear reactors will face the problem of decommissioning at the end of their operating life due to the high radioactivity of reactor components and environmental safety considerations. The Heavy Water Research Reactor (HWRR) is the first large-scale research reactor to be decommissioned in China. The second phase of HWRR decommissioning involves the main components in the reactor block, so the radiation source terms and the radioactive waste level need to be evaluated before demolition and disposal. Based on the operating history, three-dimensional geometry, materials, and other information of the HWRR, the activity of radionuclides in the main components of HWRR is calculated and analyzed, and the MCNP/ORIGEN coupling scheme is utilized for theoretical analysis. The theoretical results indicate that 14C, 54Mn, 55Fe, 60Co, 63Ni, and 152Eu are the main radioactive nuclides. The total activity of radioactive nuclides was 2.36E + 15Bq at the end of 2007, 4.27E + 13Bq at the end of 2021, and 1.83E + 13Bq at the end of 2025. Furthermore, local sampling and radiometric analyses based on the HPGe gamma-ray spectrometer are also performed to verify the theoretical results, the ratio of theoretical activity values to the measured activity of the experimental sample is within 2.5 times, so the theoretical results are conservative. According to the classification standards for radioactive waste, the inner shell, outer shell, cooling water tank, sand layer, and heavy concrete shielding layer are all low-level waste. These results and conclusions can serve as a reference for the second phase decommissioning of the HWRR and the subsequent disposal of radioactive waste.

The Application of Laser-Scanning 3D Model Reconstruction Technology for Visualizing a Decommissioning Model of the Heavy Water Research Reactor

Currently, over 100 nuclear power units globally have been in operation for more than 40 years. Hindered by the limitations of computer technology at the time, these nuclear facilities lack detailed electronic drawings. Activities such as equipment replacement and process circuit system modifications during operation result in discrepancies between paper drawings and actual conditions. Given the complexity and irreversibility of nuclear facility decommissioning activities, virtual simulation technology is often employed before the decommissioning process begins to assist in designing and validating decommissioning plans. Consequently, the creation of high-precision 3D models is crucial for subsequent decommissioning designs. Through innovatively utilizing laser-scanning 3D model reconstruction technology in the reconstruction of the model of China’s first heavy water research reactor undergoing decommissioning, this paper provides an overview of the process of laser-scanning 3D model reconstruction and its application in reconstructing the heavy water research reactor model. Using a 3D laser scanner, four decommissioning areas of the heavy water research reactor, including the reactor building, secondary water pump room, ventilation center, and low-level radioactive wastewater storage tank area, were subjected to 3D laser scanning. The acquired point cloud data from 572 scanning stations were processed using point cloud processing software for denoising, stitching, and triangulation. The triangulated model was then imported into modeling software for 3D reconstruction, ultimately establishing a digitalized model of the heavy water research reactor suitable for subsequent decommissioning simulation and design.

Application of plasma cutting in mock-up experiments for decommissioning of internals in heavy water research reactor

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.

A brief investigation of the dose field virtual simulation tools for reactor decommissioning and preliminary design for the HWRR reactor in China

The calculation and visualization of the dose field in the decommissioning of nuclear facilities is one of the important functions of the decommissioning virtual simulation system. The dose field simulation tools can provide radiation field distribution and play an important role in determining the decommissioning plan and protecting personnel during the engineering implementation process. This article investigates the development of dose field calculation and visualization in the reactor decommissioning virtual simulation systems. A preliminary technology plan suitable for the development of the decommissioning dose field calculation and visualization display programs of the first Heavy Water Research Reactor (HWRR) in China is proposed. The applicability of the selected scheme is analyzed. The functional requirement and development direction of the HWRR reactor decommissioning dose field tool are preliminarily determined. Furthermore, the reactor vessel of HWRR reactor is modeled, the dose field distribution is calculated and visualized based on the preliminary decommissioning code. This research can provide technical support for the development of the decommissioning simulation system for the first HWRR reactor in China.

Design and application of a special underwater cutting device for the decommissioning of the first HWRR reactor in China

The Heavy Water Research Reactor (HWRR) labeled 101 is the first nuclear reactor in China and the first large scale research reactor to be decommissioned in China. In the first phase of decommissioning, all spent fuel assemblies need to be removed from the spent fuel storage pool and transported to the spent fuel reprocessing plant through the RY-I transport container. The spent fuel storage pool can undergo further comprehensive cleaning and refurbishment. The HWRR reactor has operated for about fifty years, various experiment projects and several large scale upgrades were undertaken during its operation, the types and forms of spent fuel assemblies are relatively complex. Due to the deformation of the corrugated plates of some spent fuel assemblies, these spent fuel assemblies cannot be removed from the external process tubes according to the normal operation. Therefore, this paper designed a special device for the underwater cutting of the process tubes based on the structure of the process tube and spent fuel assembly, which successfully completed the underwater cutting of process tubes and the retrieval of twenty spent fuel assemblies, so all the spent fuel assemblies in spent fuel storage pool of HWRR reactor are taken out and shipped out. This device and technology provide technical reference for the underwater cutting of highly radioactive metal pipelines in other research reactors and nuclear power plants, and it can also improve the technical system and capacity for the decommissioning of the nuclear facilities in China.

Tensile Property of Irradiated LT21 Aluminum Alloy Sampled from Decommissioned Irradiation Channel of Heavy Water Research Reactor.

LT21 a type of aluminum alloy used for the irradiation channel of the first heavy water research reactor (HWRR) in China. Studying the mechanical property of irradiated LT21 aluminum under actual service conditions is essential for evaluating its application property. In this paper, tensile specimens of irradiated LT21 were manufactured from the decommissioned irradiation channel of an HWRR; then, tensile tests were carried out, and then the fracture surfaces were observed. The effect of neutron irradiation on tensile behavior and the failure mechanism was analyzed by comparing the result of irradiated and unirradiated LT21 specimens. The results show that, with the thermal neutron flux increasing to 2.38 × 1022 n/cm2, the YS gradually increased from the initial 158 MPa to 251 MPa, the UTS increased from 262 MPa to 321 MPa, and the elongation decreased from 28.8% to about 14.3%; the brittle fracture of the LT21 specimen appeared after irradiation, and the proportion of brittle fracture increased as the neutron fluence increased; the nanophase structures, with a size of less than 50 nm, were precipitated in the LT21 aluminum alloy after neutron irradiation. Transmutation Si is presumed to be the main cause of the radiation effect mechanism of LT21.

Pellet-cladding mechanical interaction analysis of heavy water fuel rods under power ramps

A numerical simulation is used to analyze the pellet-cladding mechanical interaction (PCMI) of a heavy water research reactor (RR) fuel rods under power ramps. A mechanical simulation that best reproduces the thermomechanical behavior of the pellet-cladding system is a major industrial challenge for engineers because it involves defining operating rules that are not too restrictive but that guarantee safety requirements. The majority of currently existing codes for processing thermomechanical behavior, under irradiation, of the PCMI evolution are designed for light water reactors. This is justified by the fact that most of the world nuclear power reactor park is composed of light water reactors. However, many nuclear reactors use heavy water as coolant, e.g. type CANDU reactors. In addition to the coolant thermodynamics table which is different for the two types of codes, the conditions of use in terms of pressure and temperature in the case of RRs are less severe than those in the case of nuclear power reactors (NPRs). In this study, we used several computer codes namely MCNP5, RODBURN, FEMAXI-6, and an analytical application, developed locally, to simulate the thermomechanical behavior of the Es-Salam RR fuel rods. The fission product cross-section library that was used with the RODBURN code was produced by using the ORIGEN code fed with the neutron flux spectrum of the Es-Salam research reactor. A new approach has been developed for the PCMI analysis of a heavy water reactor by using a code originally developed for light water reactors. Operating limits have been determined when the two tables of thermodynamics of coolants are different.

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.

Overcoming the uncertainty in a research reactor LOCA in level-1 PSA; Fuzzy based fault-tree/event-tree analysis

Probabilistic safety assessment (PSA) which plays a crucial role in risk evaluation is a quantitative approach intended to demonstrate how a nuclear reactor meets the safety margins as part of the licensing process. Despite PSA merits, some shortcomings associated with the final results exist. Conventional PSA uses crisp values to represent the failure probabilities of basic events. This causes a high level of uncertainty due to the inherent imprecision and vagueness of failure input data. In this paper, to tackle this imperfection, a fuzzy approach is employed with fault tree analysis and event tree analysis. Thus, instead of using the crisp values, a set of fuzzy numbers is applied as failure probabilities of basic events. Hence, in the fault tree and event tree analysis, the top events and the end-states frequencies are treated as fuzzy numbers. By introducing some fuzzy importance measures the critical components which contribute maximum to the system failure and total uncertainty are identified. As a practical example, under redesign Iranian heavy water research reactor loss of coolant accident is studied. The results show that the reactor protection system has the largest index in sequences lead to a core meltdown. In addition, the emergency core cooling system has a main role in preventing abnormal conditions.

Analysis of Design Extension Conditions in a heavy water research reactor

In this study, the core behavior following the reactor vessel lower head rupture together with the moderator system failure in a tank type heavy water research reactor, IHWRR, is analyzed through the MELCOR code. The focus is on the safety assessment of the reactor core for Design Extension Conditions and overall assessment of safety features of the reactor for any residual risk posed by severe accidents. The inherent features of IHWRR provide a broad spectrum of scenarios where the fuel does not melt, even if primary and moderator cooling are lost. Accordingly, coincident with vessel rupture, three different scenarios are considered for moderator failure: i) loss of moderator forced circulation, calandria tube rupture in ii) the upper and iii) the lower parts of the core. The obtained results showed that the vessel rupture that coincides with the tube rupture in the lower part of the core leads to the oxidation of entire Zr cladding and failure of the calandria tubes without core melt or hydrogen explosion in the containment.

Neutronic Analysis of (ThO2–EnrichedUO2) Replacement with natUO2 Fuel in a Typical Heavy Water Research Reactor

Theoretical and experimental investigations, which are exclusively carried out, reveal that thorium-based fuel could potentially be a proper alternative in nuclear fuel cycle technology. In this study, the neutronic impacts of such cycle as a consequence of natUO2 fuel substitution with an enriched uranium/thorium oxide fuel in a heavy water research reactor is presented. The obtained computational data shows that the thorium-based fuel has a less negative temperature coefficient comparing to UO2 fuel. Meantime, the thorium-based fuel loading in a typical core configuration does not drastically reduce the β as in case of UO2 fuel. Rather higher conversion factor and less transuranic production in the thorium-based fuel cycle make it favorable either to achieve higher cycle length or less dangerous and costly waste disposal.

A conceptual design for a generic heavy water research reactor to a more proliferation-resistant reactor

This paper presents a preliminary neutronic conceptual design for a heavy water research reactor core substituting low-enriched uranium fuel for natural uranium dioxide fuel; focusing on reducing the amount of plutonium in the spent fuel. In this study, two core models are proposed and investigated. These result in a smaller core configuration and consequently less plutonium inventory at the end of the cycle. Parameters such as multiplication factors, fuel burn-up, and radial power peaking factor are calculated by the MCNPX2.6 code. The results show that in converted reactor cores the mass of plutonium is decreased about 85% comparing to the main core without substantial changes in the core and fuel dimension.

Reduction of Plutonium Production in Heavy Water Research Reactor: A Feasibility Study through Neutronic Analysis Using MCNPX2.6 and CINDER90 Codes

Reduction of Plutonium Production in Heavy Water Research Reactor: A Feasibility Study through Neutronic Analysis Using MCNPX2.6 and CINDER90 Codes

Safety analysis of reactivity insertion accidents in a heavy water nuclear research reactor core using coupled 3D neutron kinetics thermal-hydraulic system code technique

Nuclear power plant Safety analysis using coupled 3D neutron kinetics/thermal-hydraulic codes technique is increasingly used nowadays. Actually, the use of this technique allows getting less conservatism and more realistic simulations of the physical phenomena. The challenge today is oriented toward the application of this technique to the operating conditions of nuclear research reactors. In the current study, a three-Dimensional Neutron Kinetics and best estimate Thermal-Hydraulic model based upon the coupled PARCS/RELAP5 codes has been developed and applied for a heavy water research reactor. The objective is to perform safety analysis related to design accidents of this reactor types. In the current study two positive reactivity insertion transients are considered, SCRAM protected and self-limiting power excursion cases. The results of the steady state calculations were compared with results obtained from conventional diffusion codes, while transient calculations were assessed using the point kinetic model of the RELAP5 code. Through this study, the applicability and the suitability of using the coupled code technique with respect to the classical models are emphasized and discussed.

Comparison of neutronic behavior of UO 2 , (Th- 233 U)O 2 and (Th- 235 U)O 2 fuels in a typical heavy water reactor

The research carried out on thorium-based fuels indicates that these fuels can be considered as economic alternatives with improved physical properties and proliferation resistance issues. In the current study, neutronic assessment of UO₂ in comparison with two (Th- 233 U)O₂, and (Th- 235 U)O₂ thorium-based fuel loads in a heavy water research reactor has been proposed. The obtained computational data showed both thorium-based fuels caused less negative temperature reactivity coefficients for the modeled research reactor in comparison with UO₂ fuel loading. By contrast, 233U-containing thorium-based fuel and 235 U-containing thorium-based fuel loadings in the thermal core did not drastically reduce the effective delayed neutron fractions and delayed neutron fractions compared to UO₂ fuel. A provided higher conversion factor and lower transuranic production in the research core fed by the thorium-based fuels make the fuel favorable in achieving higher cycle length and less dangerous and costly nuclear disposals.

Two Methods for Converting a Heavy-Water Research Reactor to Use Low-Enriched-Uranium Fuel to Improve Proliferation Resistance After Startup

AbstractThis article demonstrates the feasibility of converting a heavy-water research reactor from natural to low-enriched uranium in order to slow the production of weapon-usable plutonium, even if the core cannot be physically reconfigured. The analysis was performed for Iran’s IR-40 reactor at Arak in support of negotiations with Iran, but the methods have application to future reactors that present similar nonproliferation challenges. Two methods are considered, and both retain identical power, thermal-hydraulic, and safety profiles as the original reactor design. The conversion options can be implemented at any time during the reactor’s life. The two methods have competing effects on achievable burnup, and they can be combined to produce an optimized core that matches both the fresh-core reactivity and maximum burnup of the original reactor. For the IR-40 example, the optimized design produces weapon-grade plutonium at only about 19% of the rate of the unmodified reactor for the same power level. Ad...

Converting the Iranian Heavy Water Reactor IR-40 to a More Proliferation-Resistant Reactor

This article assesses the feasibility and benefits of converting the Iranian heavy water research reactor, IR-40, from using natural uranium to low-enriched uranium fuel. Based on neutronics calculations for a detailed model of the two reactor configurations, a conversion would result in a smaller core with a subsequent reduction and degradation of plutonium production. It is argued that the proposed conversion will provide Iran with a research reactor that is better suited for scientific experiments and radioisotope production than in its original configuration. It is proposed to introduce the converted IR-40's fuel consumption requirements as a natural cap for Iran's future enrichment efforts.

Evaluate the radioactivity along the central thimble hole of a decommissioned heavy water research reactor using TLD approach

The radioactivity along the central thimble hole of a decommissioned heavy water research reactor, TRR, was evaluated using TLD approach. The decay radionuclide was verified to be Co-60. The dose along the TRR central thimble hole was detected and revised by performing an unfolding analysis. The revised data reduced to 70–90% of the original data (for example, the maximum dose rate was reduced from 6447 to 4831mSv/h,) and were more reliable.

Neutronic Analysis of Generic Heavy Water Research Reactor Core Parameters to Use Standard Hydride Fuel

This research presents neutronic calculations for heavy water research reactor core substituting hydride fuel for uranium dioxide fuel. The aim of this research is feasibility analysis of reactor utilization with its original design using a new proposed fuel and changing the coolant and moderator circuit to light water. The required group constants for the CITATION code will be calculated using WIMSD-4 code. Neutronic calculations such as multiplication factors, radial and axial power peaking factor and fuel burn-up calculations are carried out by the CITATION code.

HAZOP-study on heavy water research reactor primary cooling system

Abstract By knowledge-based Hazard and Operability (HAZOP) technique, equipment malfunction and deficiencies in the primary cooling system of the generic heavy water research reactor are studied. This technique is used to identify the representative accident scenarios. The related Process Flow Drawing (PFD) is prepared as our study database for this plant. Since this facility is in the design stage, applying the results of HAZOP-study to PFD improves the safety of the plant.