Heavy Water Moderator Research Articles

Comparison of the Reactivity Effects Calculated by DRAGON and Serpent for a PHWR 37-Element Fuel Bundle

Deterministic and Monte Carlo methods are regularly employed to conduct lattice calculations. Monte Carlo methods can effectively model a large range of complex geometries and, compared to deterministic methods, they have the major advantage of reducing systematic errors and are computationally effective when integral quantities such as effective multiplication factor or reactivity are calculated. In contrast, deterministic methods do introduce discretization approximations but usually require shorter computation times than Monte Carlo methods when detailed flux and reaction-rate solutions are sought. This work compares the results of the deterministic code DRAGON to the Monte Carlo code Serpent in the calculation of the reactivity effects for a pressurized heavy water reactor (PHWR) lattice cell containing a 37-element, natural uranium fuel bundle with heavy water coolant and moderator. The reactivity effects are determined for changes to the coolant, moderator, and fuel temperatures and to the coolant and moderator densities for zero-burnup, mid-burnup [3750 MWd/t(U)] and discharge burnup [7500 MWd/t(U)] fuel. It is found that the overall trend in the reactivity effects calculated using DRAGON match those calculated using Serpent for the burnup cases considered. However, differences that exceed the amount attributable to statistical error have been found for some reactivity effects, particularly for perturbations to coolant and moderator density and fuel temperature.

INVESTIGATION OF REACTOR PHYSICS PHENOMENA IN THE CANADIAN PRESSURE TUBE SUPERCRITICAL-WATER REACTOR

The Canadian pressure-tube supercritical-water-cooled reactor is an advanced Generation IV reactor concept that uses water above its thermodynamic critical pressure as its coolant. The higher operating pressure dictates changes in the design and configuration of the reactor core and fuel assemblies as compared with existing CANDU designs. In addition, the reference Canadian design considers a plutonium-driven thorium fuel with high-pressure light-water coolant and a separate low-pressure heavy-water moderator. The salient features of this core design include a vertically orientated re-entrant channel where the coolant passes from an inlet plenum above the core downwards through the internal annulus flow tubes in each fuel assembly, then through a 180° bend, then upwards through the fuel region of the assembly where it absorbs thermal energy, and finally into the outlet plenum located above the core. Given the multiple flow paths through a fuel assembly, the significant coolant property variations along the fuel channel caused by transitions through the pseudo-critical temperature, the external low-pressure moderator, and the Pu–Th fuel composition, the lattice physics phenomena are significantly different from those in conventional CANDU power plants. In this paper, the changes in lattice physics phenomena are identified and analyzed through sensitivity and uncertainty analyses using the lattice physics modules in standardized computer analysis for licensing evaluation.

CONCEPTUAL PLANT LAYOUT OF THE CANADIAN GENERATION IV SUPERCRITICAL WATER-COOLED REACTOR

Canada is developing a 1200 MWe supercritical water-cooled reactor (SCWR), which has evolved from the well-established pressure-tube type CANDU®1 reactor. This SCWR reactor concept, which is often referred to as the Canadian SCWR, uses supercritical water as a coolant, has a low-pressure heavy water moderator, and a direct cycle for power production. The reactor concept incorporates advanced safety features, such as passive emergency core cooling, long-term decay heat rejection to the environment, and fuel melt prevention via passive moderator cooling. To support these features, major systems and major components are identified and laid out inside the reactor buildings.The Canadian SCWR uses 2 nested reactor buildings for 2 separate functions: (i) a leak-tight containment building to house all safety-related systems and (ii) a shield building to protect critical components against external threats, such as airplane crashes, missile attacks, tornados, and flooding. Two redundant and independent passive saf...

Poison injection in AHWR calandria: Flow pattern and mixing characteristics

In Advanced Heavy Water Reactor (AHWR), two active, independent, functionally diverse, fast acting shut down systems are used. Shut Down System-1 (SDS-1) consists of mechanical shut off rods and Shut Down System-2 (SDS-2) based on liquid poison injection into the moderator. Gadolinium nitrate solution, which acts as neutronic poison, is injected in heavy water moderator through a set of nozzles located inside calandria vessel. This poison injection into moderator inside calandria is expected to take place in a very short time to shut down the reactor. As SDS-2 is critical for the safety of plant, detailed knowledge of poison distribution in calandria is required.CFD analysis using OpenFOAM has been carried out for high pressure poison injection of Shut Down System-2 of AHWR for various design options. Validation of CFD model is performed by experimental measurements. Experiments are performed for impulsively started poison jet in a 1/3.3 scaled and full-scale experimental facilities with array of tubes. In particular, the growth of poison jet is measured experimentally and CFD simulations are performed. An excellent agreement is found between CFD simulations and experimental data. CFD simulation shows that the poison covers a major portion of active core in 2s time. In addition to this, effect of moderator flow and heat generation in moderator on poison distribution has also been studied.

Experimental and CFD simulations of fluid flow and temperature distribution in a natural circulation driven Passive Moderator Cooling System of an advanced nuclear reactor

The Passive Moderator Cooling System (PMCS) of the Advanced Heavy Water Reactor is designed to remove heat from the moderator passively in case of an extended station black out condition (SBO). The hot heavy-water moderator inside the Calandria rises upward due to buoyancy, cooled in a shell and tube heat exchanger (located within the loop) and returns back to the Calandria, completing a natural circulation loop. The heat exchanger, in turn, is cooled by water from a Gravity Driven Water Pool (GDWP); forming a second loop. This coupled natural circulation loop system provides sufficient cooling to prevent the increase of moderator temperature inside the Calandria vessel beyond safe limits during SBO.The feasibility of such a system should be assessed before it is implemented in the reactor. Thus, a scaled test facility was set up to simulate the thermal hydraulic characteristics of the PMCS. A set of time varying power experiments were performed, which capture the flow initiation from rest phenomena and the multidimensional natural convection flow in a coupled natural circulation system. Next, the experimental geometry was simulated using the three-dimensional computational fluid dynamics code (OpenFoam 2.2.0), which predicted temperature and flow distribution inside the system. The CFD results agree well with the experimental data within ±6%. The flow initiation phenomena shows that the initial flow generated recirculates within the Calandria, after ~900s the flow is able to come out of the Calandria. The experiments show a time lag between the starting of the primary and secondary loop circulation. This work demonstrates the experimental and computational capability to understand and design an effective Passive Moderator Cooling System.

Analysis of neutronics benchmarks for the utilization of mixed oxide fuel in light water reactor using DRAGON code

Advances in reactor physics have led to the development of new computational technologies and upgraded cross-section libraries so as to produce an accurate approximation to the true solution for the problem. Thus it is necessary to revisit the benchmark problems with the advanced computational code system and upgraded cross-section libraries to see how far they are in agreement with the earlier reported values. Present study is one such analysis with the DRAGON code employing advanced self shielding models like USS and 172 energy group ‘JEFF3.1’ cross-section library in DRAGLIB format. Although DRAGON code has already demonstrated its capability for heavy water moderator systems, it is now tested for light water reactor (LWR) and fast reactor systems. As a part of validation of DRAGON for LWR, a VVER computational benchmark titled “Neutronics Benchmarks for the Utilization of Mixed-Oxide Fuel-Volume 3” submitted by the Russian Federation has been taken up. Presently, pincell and assembly calculations are carried out considering variation in fuel temperature (both fresh and spent), moderator temperatures and boron content in the moderator. Various parameters such as infinite neutron multiplication (k∞) factor, one group integrated flux, few group homogenized cross-sections (absorption, nu-fission) and reaction rates (absorption, nu-fission) of individual isotopic nuclides are calculated for different reactor states. Comparisons of results are made with the reported Monte Carlo (MCU) values of the benchmark. Maximum deviation of 1.8% in k∞ is observed for variants with spent fuel and for the states with control rod whereas all the other results are in par with the results reported in the benchmark document. The few and multi-group macroscopic cross-sections and flux of all the nuclides also compare well with the benchmark results except for the 11B macroscopic absorption cross section, which is further compared with the XNWLUP software. Inter-comparison of results with the generalized self-shielding model SHI of DRAGON code employing the traditional WIMSD formatted 172 group cross-section library has also been made to highlight the improvements made in computational schemes and cross-section library format.

Coupled 3D neutron kinetics and thermalhydraulic characteristics of the Canadian supercritical water reactor

The Canadian Supercritical Water-cooled Reactor concept, as an evolution of the CANada Deuterium Uranium (CANDU) reactor, includes both pressure tubes and a low temperature heavy water moderator. The current Pressure Tube type SCWR (PT-SCWR) concept features 64-element fuel assemblies placed within High Efficiency Re-entrant Channels (HERCs) that connect to core inlet and outlet plena. Among current SCWR concepts the PT-SCWR is unique in that the HERC separates multiple coolant and moderator regions, giving rise to coupled neutronic-thermalhydraulic feedbacks beyond those present in CANDU or contemporary Light Water Reactors. The objective of this work was thus to model the coupled neutronic-thermal hydraulic properties of the PT-SCWR to establish the impact of these multiple regions on the core's transient behavior. To that end, the features of the PT-SCWR were first modeled with the neutron transport code DRAGON to create a database of homogenized and condensed cross-sections and thermalhydraulic feedback coefficients. These were used as input to a core-level neutron diffusion model created with the code DONJON. The behavior of the primary heat transport system was modeled with the thermalhydraulic system code CATHENA. A procedure was developed to couple the outputs of DONJON and CATHENA, facilitating three-dimensional spatial neutron kinetics and coupled thermalhydraulic analysis of the PT-SCWR core. Several postulated transients were initiated within the coupled model by changing the core inlet and outlet boundary conditions. Decreasing coolant density around the fuel was demonstrated to produce positive power excursions (i.e., the coolant void reactivity around the fuel was positive), but such power transients were found to be inherently self-terminating as low density coolant inevitably reaches other parts of the HERC geometry (where the void reactivity is highly negative). Nevertheless, the observed power excursions potentially demonstrate the need for fast-acting shutdown system intervention, similar to CANDU designs.

Various Design Aspects of the Canadian Supercritical Water-Cooled Reactor Core

The Canadian Supercritical Water-Cooled Reactor (SCWR) is a 1200 MW(e) channel-type nuclear reactor. The reactor core includes 336 vertical pressurized fuel channels immersed in a low-pressure heavy water moderator and calandria vessel containment. The supercritical water (SCW) coolant flows into the fuel channels through a common inlet plenum and exits through a common outlet header. One of the main features of the Canadian SCWR concept is the high-pressure (25 MPa) and high-temperature (350°C at the inlet, 625°C at the outlet) operating conditions that result in an estimated thermal efficiency of 48%. This is significantly higher than the thermal efficiency of the present light water reactors, which is about 33%. This paper presents a description of the Canadian SCWR core design concept; various numerical analyses performed to understand the temperature, flow, and stress distributions of various core components; and how the analyses results provided input for improved concept development.

CFD simulations of moderator flow inside Calandria of the Passive Moderator Cooling System of an advanced reactor

Passive systems are being examined for the future Advanced Nuclear Reactor designs. One of such concepts is the Passive Moderator Cooling System (PMCS), which is designed to remove heat from the moderator in the Calandria vessel passively in case of an extended Station Black Out condition. The heated heavy-water moderator (due to heat transferred from the Main Heat Transport System (MHTS) and thermalization of neutrons and gamma from radioactive decay of fuel) rises upward due to buoyancy, gets cooled down in a heat exchanger and returns back to Calandria, completing a natural circulation loop. The natural circulation should provide sufficient cooling to prevent the increase of moderator temperature and pressure beyond safe limits. In an earlier study, a full-scale 1D transient simulation was performed for the reactor including the MHTS and the PMCS, in the event of a station blackout scenario (Kumar et al., 2013). The results indicate that the systems remain within the safe limits for 7 days. However, the flow inside a geometry like Calandria is quite complex due to its large size and inner complexities of dense fuel channel matrix, which was simplified as a 1D pipe flow in the aforesaid analysis. In the current work, CFD simulations are performed to study the temperature distributions and flow distribution of moderator inside the Calandria vessel using a three-dimensional CFD code, OpenFoam 2.2.0. First, a set of steady state simulation was carried out for a band of inlet mass flow rates, which gives the minimum mass flow rate required for removing the maximum heat load, by virtue of prediction of hot spots inside the Calandria. Next, a pseudo transient simulation was carried out by taking the power and mass flow rate at different time instants to the Calandria from the 1D analysis, as an input to the 3D steady state simulations. The results indicate that temperature hotspots are located at the top center of the fuel channel matrix depending on the operating regime. Furthermore, it was observed that for a period of 7 days the temperatures are well within the safe limits.

Variation of Burnable Neutron Absorbers in a Heavy Water–Moderated Fuel Lattice: A Potential to Improve CANDU Reactor Operating Margins

A CANDU lattice cell has been modeled using the Los Alamos National Laboratory’s MCNP 6 code and Atomic Energy of Canada Limited’s WIMS-AECL 3.1. Models for the CANDU 37-element fuel bundle have included a CANLUB coating, as a carrier for the neutron absorbers. The objective is to improve CANDU reactor operating margins by adding small amounts (~1 g) of neutron absorbers to each fuel element.For CANDU natural uranium fuel bundle design, the results indicate that (a) the fueling transient (due to the xenon-free effect) could be significantly reduced using gadolinium oxide (Gd2O3), with no significant impact on fuel burnup, and (b) the reactivity peak (due to plutonium production) could be reduced using europium oxide (Eu2O3), with minimal impact on fuel burnup. An appropriate mixture of Gd2O3 and Eu2O3 that will improve operation and safety margins while having a minimal impact on fuel burnup is determined.Reactivity and power calculations for various mixtures of Gd2O3 and Eu2O3 are reported here. It is concluded that ~180 mg Gd2O3 and ~1000 mg Eu2O3 (~4.9 ×10−3 wt% per bundle) are sufficient to suppress the refueling transient and lower the axial plutonium peak, with a 0.27% burnup penalty (which is a small impact).Fuel safety and performance are always important topics for a nuclear utility. This approach of a relatively simple application of burnable poisons to existing CANDU natural uranium fuel design offers the benefits of improving fuel utilization and safety margins.

Development of Tritium Technologies at KAERI

Tritium is formed by neutrons captured from deuterium. If left to accumulate, tritium oxide will become a hazard to the operating staff and public. The primary purpose of a Tritium Removal Facility (TRF) is to reduce tritium concentration in a heavy water moderator. In Korea, operation of a TRF commenced at the Wolsong Nuclear Power Site on July 26th, 2007. Nowadays, KAERI is developing a Very High Temperature Gas Cooled Reactor (VHTR). We have developed a tritium behavior analysis code for the VHTR. We also developed analytical methods for the measurement of food stuffs. Korea shared in the construction of the ITER fuel cycle plant with the EU, Japan, and the US, and is responsible for the supply of an SDS (Tritium Storage and Delivery System). We present the recent progress in the development of tritium storage technology, and safety features of the related system. KAERI has been developing tritium technologies related to the Wolsong TRF, HANARO, VHTR, and nuclear fusion fuel systems. We thus present details on the recent development progress of these tritium systems.

Replacement of Tritiated Water from Irradiated Fuel Storage Bay

Recently, AECL developed a novel method to reduce tritium emissions (to groundwater) and personnel doses at the NRU (National Research Universal) reactor irradiated fuel storage bay (also known as rod or spent fuel bay) through a water swap process. The light water in the fuel bay had built up tritium that had been transferred from the heavy water moderator through normal fuel transfers. A demonstration of the method was done that involved Computational Fluid Dynamics (CFD) modeling of the swap process and a test program that showed excellent agreement with model prediction for the effective removal of almost all the tritium with a minimal water volume. Building on the successful demonstration, AECL fabricated, installed, commissioned and operated a full-scale system to perform a water swap. This paper describes the results of the successful full-scale water swap operation, which achieved a tritium removal efficiency of ~96%

A PRELIMINARY CATHENA THERMALHYDRAULIC MODEL OF THE CANADIAN SCWR FOR SAFETY ANALYSIS

The supercritical water-cooled reactor (SCWR) is one of six reactor concepts under development in the Generation-IV International Forum (GIF). As a member of GIF, Canada is developing a pressure-tube type SCWR, which has the potential to fulfill all major GIF goals on enhanced safety, sustainability, economics, and proliferation resistance. The system thermalhydraulics code CATHENA will be used in the safety analyses for the Canadian SCWR. Based on the current conceptual design of the Canadian SCWR, a CATHENA idealization has been developed. This model includes all 336 fuel channels with a detailed model of heat transfer in the reactor core. Also modeled are the main pumps, inlet plenum, outlet plenum, turbines, and heavy water moderator. In this paper, the CATHENA idealization of the Canadian SCWR conceptual design is described. Simulation results for steady-state normal operations are also presented for the current Canadian SCWR conceptual design.

Optimization of 54-element pressure tube SCWR bundle through neutronics/thermalhydraulics coupling analysis

The radial power distribution is uneven in the Pressure Tube type SuperCritical Water-cooled Reactor (PT-SCWR) bundle because of the high fuel enrichment, the separation of light water coolant and heavy water moderator. The uneven profile will usually result in higher cladding temperature which cannot meet the design limit for SCWR. In order to optimize bundle design, the analysis of coupled neutronics/thermal-hydraulics was carried out.The neutronics code WIMS-AECL and the subchannel code ATHAS are selected to conduct the coupled neutronics/thermal-hydraulics analysis for the PT-SCWR 54-elements bundle design. The optimization was carried out through (1) adopting staggered radial uranium enrichment profile and (2) optimizing the pitch-circle radiuses of element rings.The results showed that the obtained optimized bundle can meet the demand of limiting criteria for peak cladding temperature and negative coolant temperature coefficient.

Neutronic behavior of reactor moderated by mixtures of light and heavy waters at different ratios

A Monte-Carlo parametric study was carried out to investigate some nuclear properties of LR-0 reactor when it is moderated by mixtures of light and heavy water at molecular ratios ranging from 0% up to 100% D2O at increments of +10% in D2O. Neutron energy spectrum, fission rates, non-fission neutron capture rates, neutron absorption in poison material, and average fission energy yield from 235U and 238U components of the UO2 fuel were tallied at different regions in the reactor at the 11 heavy water percentage moderators being studied. It was found that the changes in all the tallied parameters with heavy water percentage in moderator were not monotonic, and the nuclear properties of the reactor changed dramatically between 90% and 100% heavy water moderators. The most prominent changes were in neutron energy spectrum, neutrons leaking out from the reactor, neutrons absorption in moderator, and non-fission neutron capture in 238U. These changes were not the same in all regions of the reactor.

Salt-Cooled Modular Innovative Thorium Heavy Water-Moderated Reactor System

This paper describes a new reactor concept: the Salt-cooled Modular Innovative THorium HEavy water-moderated Reactor System (SMITHERS), which addresses the goals of (a) evolving deployment needs, (b) increasing overall fuel burnup, (c) reducing proliferation risk, and (d) providing high-efficiency power generation. The reactor is modular and thus scalable from a few to hundreds of megawatts(thermal). The concept further burns used fuel from light water reactors (LWRs) without aqueous separations, reducing costs and proliferation pathways relative to current reprocessing plants. The additional burning of LWR fuel reduces proliferation risk by reducing global inventories of plutonium from used fuel in a way that does not isolate weapons-useable material and that increases the amount of power produced per ton of mined uranium. Improved fuel utilization through the potential use of thorium provides cost benefits by increasing neutron economy and enabling operation at higher efficiencies. Neutron economy is increased by using the lower neutron energies associated with large quantities of heavy water moderation and/or thorium for innovative reactor control and constant long-term power generation (i.e., sustainability). Finally, the proposed reactor also generates high-temperature coolant discharge in the form of liquid salt without coolant pressurization for external process heat applications such as oil extraction. Salt offers significant improvement over existing coolants such as light water and heavy water, which require pressurization to operate at high temperatures, adding to the cost and complexity of reactor operation. SMITHERS designs discussed in this paper either burned a full core of used fuel, ThO2 with 1.2 wt% PuO2 or other fissile material, or a combination of the two.

Analysis of Interaction of Calandria Tubes with the End Shields in Advanced Heavy Water Reactor (AHWR)

Advanced Heavy Water Reactor (AHWR) is a 300 MWe vertical pressure tube type, boiling light water cooled and heavy water moderated reactor. It consists of 452 coolant channels wherein the heat of nuclear reaction is used to boil the light water coolant to generate steam. Typical coolant channel assembly of AHWR consists of a pressure tube of cold worked Zr- 2.5Nb alloy extended by stainless steel end-fittings, attached at its ends. A Calandria tube of annealed Zircaloy-4 alloy separates the hot pressure tube from the cold heavy water moderator. The Calandria tube is attached at its ends to the Calandria side tube-sheets of both the bottom and top end shields. Thermal expansion, irradiation enhanced axial creep and growth of Calandria tubes cause them to elongate. These elongations induce compressive stresses in Calandria tubes and also load the tube-sheets of the end shields. Stresses also get induced in the annular plate, main-shell and sub-shells of the Calandria vessel. Nature of interaction and types of stresses induced in these components will depend upon the support conditions of both the end shields, creep and growth rates of Calandria tubes. This has been studied for different combinations of fixed and floating types of support conditions of both the top and the bottom end shields. It has been observed that compared to the fixed support condition of both the end shields, stresses induced in the different components reduce to half when either of the end shields support condition is kept floating or on some flexible support.

Developments Toward Mechanistic Modeling of Critical Heat Flux on a CANDU Calandria Tube Following Pressure Tube Contact

Heavy water moderator surrounding each fuel channel is one of the important safety features in CANDU reactors since it provides an in situ passive heat sink for the fuel in situations where other engineered means of heat removal from fuel channels have failed. In a critical-break loss-of-coolant-accident scenario, fuel cooling becomes severely degraded because of rapid flow reduction in the affected flow pass of the heat transport system. This can result in pressure tubes (PTs) experiencing significant heatup during early stages of the accident when coolant pressure is still high, thereby causing uniform thermal creep strain (ballooning) of the PT in contact with its calandria tube (CT). The contact of the hot PT with the CT causes rapid redistribution of stored heat from the PT to the CT and a large heat flux spike from the CT to the moderator fluid. For conditions where subcooling of the moderator fluid is low, this heat flux spike can cause dryout of the CT. This can detrimentally affect channel integrity if the CT postdryout temperature becomes sufficiently high to result in continued thermal creep strain deformation of both the PT and the CT. The focus of this work is to develop a mechanistic model to predict critical heat flux (CHF) on the CT surface following a contact with its PT. A COMSOL multiphysics model using a two-dimensional transient fluid-thermal analysis of the CT surface undergoing heatup is used to predict the flow and temperature profiles on the CT surface. A mechanistic CHF model is to be proposed based on a concept of wall dry patch formation, prevention of rewetting, and subsequent dry patch spreading.

Commercial utilization of weapon grade plutonium as TRISO fuel in conventional CANDU reactors

Large quantities of weapon grade (WG) plutonium have been accumulated in the nuclear warheads. Plutonium and heavy water moderator can give a good combination with respect to neutron economy. TRISO type fuel can withstand very high fuel burn up levels. The paper investigates the prospects of utilization of TRISO fuel made of WG-plutonium in CANDU reactors.Three different fuel compositions have been investigated: (1): 90% ThC+10% PuC, (2): 70% ThC+30% PuC and (3): 50% ThC+50% PuC. The temporal variation of the criticality k∞ and the burn-up values of the reactor have been calculated by full power operation up to 17years. Calculated startup criticalities for these fuel modes are k∞,0=1.6403, 1.7228 and 1.7662, respectively.Attainable burn up values and reactor operation times without new fuel charge will be 94700, 265000 and 425000MW.D/MT and along with continuous operation periods of ∼3.5, 10 and 17years, respectively, for the corresponding modes. These high burn ups would reduce fuel fabrication costs and nuclear waste mass for final disposal per unit energy drastically.

LBLOCA in CANDU-NG cooled by light water

The purpose of this work is to develop methodologies for the evaluation of LBLOCA in CANDU-NG reactors with the codes DONJON and DRAGON. CANDU-NG reactor differ from traditional CANDU reactors in being cooled by light water, using enriched fuel and burnable poisons, having significantly lesser quantity of heavy water moderator. The evaluation shows that methodology developed for CANDU-NG LBLOCA properly detects positive reactivity introduced in the core by initial voiding in checkerboard pattern, peaking at 143pcm. Such reactivity quickly becomes negative, however, bottoming at −804pcm and the reactor shuts down by itself without the intervention of any engineered system.