Light Water-cooled Reactor Research Articles

Experience feedback on operation of low-temperature vacuum evaporation-based desalination plant integrated with CIRUS

CIRUS is a 40 MWth, heavy water-moderated, light water-cooled and natural uranium-fuelled research reactor located at Trombay, Mumbai. The reactor has been in operation since 1960 with an average availability factor of about 70% for about three decades. During the early 1990s, the reactor exhibited signs of ageing. It was refurbished and put back into operation at high power in 2003. Bhabha Atomic Research Centre (BARC) has been engaged in R&D activities related to desalination. Using low-temperature vacuum evaporation technology, the centre has been studying the possibility of using the waste heat of nuclear reactors for seawater desalination. Based on this know-how, a 30 cu m per day (30 tonnes/day) pilot plant has been designed. With the aim of demonstrating the utilisation of waste heat from research reactor, it was decided to integrate the unit to CIRUS during refurbishing outage of the reactor. The work involved the design, installation and commissioning of a set-up to transfer heat from the primary coolant of CIRUS to the desalination unit through an intermediate demineralised water circuit. The unit has been operated at its rated capacity and its performance has been satisfactory. The product water has been utilised to augment the demineralised water inventory of the primary coolant system. This paper highlights the experience gained during the installation, commissioning and operation of the desalination unit.

Subchannel analysis of supercritical light water-cooled fast reactor assembly

Subchannel analyses have been carried out for supercritical water-cooled fast reactor fuel assembly. Peak cladding surface temperature difference arising from subchannel heterogeneities have been calculated by using the improved subchannel analysis code STARS and was evaluated to be about 18.5 °C. Several suggestions have been also made for reducing the PCST difference arising from channel heterogeneity. Influences of local power peaking on deflection of cladding surface temperature are explained with pin power distribution taken from core depletion calculation in this paper. Maximum cladding surface temperature at nominal condition is evaluated to be 645.3 °C over the cycle. Statistical thermal design uncertainty associated with PCST calculation is evaluated by Monte-Carlo sampling technique combined with subchannel analysis code. Maximum statistical design uncertainty of PCST is calculated to be 31 °C and is in a good agreement with that from RTDP method. Influence of downward flow in seed region on system sensitivity is investigated by improved Monte-Carlo thermal design procedure. Limiting thermal condition of MCST is 681 °C (650 °C of nominal + 31 °C) within 95/95 limit for SWFR.

Thermo-mechanical analysis of supercritical pressure light water-cooled fast reactor fuel rod by FEMAXI-6 code

Thermo-mechanical behaviors of supercritical pressure light water cooled fast reactor (SWFR) fuel rod and cladding have been investigated by FEMAXI-6 (Ver.1) code with high enriched MOX fuel at elevated operating condition of high coolant system pressure (25 MPa) and high temperature (500 °C in core average outlet temperature). Fuel rod failure modes and associated fuel rod design criteria that is expected to be limiting in SWFR operating condition have been investigated in this fuel rod design study. Fuel centerline temperature is evaluated to be 1853 °C and fission gas release fraction is about 45% including helium production. Cumulative damage fraction is evaluated by linear life fraction rule with time-to-rupture correlation of advanced austenitic stainless steel. In a viewpoint of mechanical strength of fuel cladding against creep rupture and cladding collapse at high operation temperature, currently available stainless steels or being developed has a potential for application to SWFR. Admissible design range in terms of initial gas plenum pressure and its volume ratio are suggested for fuel rod design The stress ranges suggested by this study could be used as a preliminary target value of cladding material development for SWFR application.

Control of a High Temperature Supercritical Pressure Light Water Cooled and Moderated Reactor with Water Rods

The plant system of a supercritical pressure light water reactor (SCR) is once-through direct cycle. The whole coolant from the feedwater pumps is driven to the turbines. The core flow rate is less than 1/7 of that of a boiling water reactor. In the present design of the high temperature thermal reactor (SCLWR-H), the fuel assemblies contain many water rods in which the coolant flows downward. The stepwise responses of the SCLWR-H are analyzed against perturbations without a control system. Based on these analyses, a control system of the SCLWR-H is designed. The pressure is controlled by the turbine control valves. The main steam temperature is controlled by the feedwater pumps. The reactor power is controlled by the control rods. The control parameters are optimized by the test calculations to satisfy the criteria of both fast convergence and stability. The reactor is controlled stably with the designed control systems against various perturbations, such as setpoint change of the pressure, the main steam temperature and the core power, decrease in the feedwater temperature, and decrease in the feedwater flow rate.

Control of a High Temperature Supercritical Pressure Light Water Cooled and Moderated Reactor with Water Rods

The plant system of a supercritical pressure light water reactor (SCR) is once-through direct cycle. The whole coolant from the feedwater pumps is driven to the turbines. The core flow rate is less than 1/7 of that of a boiling water reactor. In the present design of the high temperature thermal reactor (SCLWR-H), the fuel assemblies contain many water rods in which the coolant flows downward. The stepwise responses of the SCLWR-H are analyzed against perturbations without a control system. Based on these analyses, a control system of the SCLWR-H is designed. The pressure is controlled by the turbine control valves. The main steam temperature is controlled by the feedwater pumps. The reactor power is controlled by the control rods. The control parameters are optimized by the test calculations to satisfy the criteria of both fast convergence and stability. The reactor is controlled stably with the designed control systems against various perturbations, such as setpoint change of the pressure, the main steam temperature and the core power, decrease in the feedwater temperature, and decrease in the feedwater flow rate.

ICONE11-36186 INTEGRATED LEAK RATE TESTING OF CONTAINMENTS-A PERSPECTIVE

Appendix J of 10 CFR 50,A Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, @ sets the testing requirements for preoperational and periodic verification of the leak-tight integrity of the primary reactor containment, including systems and components which penetrate containment of light water-cooled power reactors, and establishes the acceptance criteria for such tests. The purposes of the tests are to assure that leakage through the primary reactor containment and systems and components penetrating primary containment would not exceed allowable leakage rate values as specified in the plants=technical specifications. The current Appendix J requirements provide two options for performing the tests. Option A (fully deterministic) requires that after the preoperational leakage rate tests, a set of three integrated leak rate tests (ILRT-termed as Type A tests) shall be performed, at approximately equal intervals during each 10-year service period. Option B does not provide a quantitative requirement for scheduling the periodic Type A tests. However, NEI 94-01, A Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50,Appendix J, @ and NRC Regulatory Guide (RG) 1.163, A Performance-Based Containment Leak Test Program, @ provide guidelines for determining the frequencies of preoperational and periodic leak rate tests using a performance based approach. The Option B requirements and NEI 94-01 guidance are based on NUREG-1493, A Performance Based Containment Leak-Test Program. @ Based on the information provided by the operating data and their risk significance, the study calculated the change in risk (in person-rem) to public for the 15 alternatives considered in the study. However, recognizing the non-sensitivity of risk to change in Type A leak rate testing frequencies, as depicted in NUREG-1493,a number of licensees are proposing changes to their Type A test frequencies using NRC risk-informed guidance in RG 1.174, A An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis. The paper discusses certain key deterministic aspects of the risk-informed decision for the plant-specific changes in the ILRT frequencies.

A Methodology for Selecting High Thermal-Hydraulic Performance Fuel Configurations for Tightly Packed Epithermal Core Designs

Cylindrical fuel pins with wires are the design of choice for tightly packed fuel arrays. However, it is important to investigate novel fuel configurations in order to increase the thermal margins. Hence, new fuel designs have been studied for the epithermal option of the light water-cooled IRIS core. These designs are also of potential use in other tightly packed, epithermal advanced core designs.First, design equations have been used to determine number, height, and size of the principal features (clad, gap, fuel cross-sectional area) of the novel fuel configurations under investigation. Then, performance indices have been introduced to relate fuel geometrical characteristics to selected thermal-hydraulic parameters, such as pressure drop, critical heat flux (CHF), fuel centerline temperature, and clad surface temperature and stress distribution. Finally, variously shaped fuel configurations, including cylindrical, triangular, square, and hexagonal, have been ranked according to the performance indicators.The hexagonal fuel pins, both twisted and straight, proved to be good solutions for the epithermal tight core of the light water-cooled IRIS reactor, with performances comparable to those of the cylindrical fuel with wires. In particular, for water-to-fuel ratios ~0.33, the twisted hexagonal shape is the preferable design with a reduction of the total pressure drop by 16% and an increase of the CHF margin by 200%, compared to the traditional cylindrical pins with grids. Furthermore, the straight hexagonal shape allows flatter subchannel velocity profiles, wall shear stress, and wall temperature distributions. However, geometric constraints unfortunately do not allow application of the twisted hexagonal shape for smaller water-to-fuel ratios, which is a design regime of more favorable epithermal neutronics performance. In this regime, the cylindrical pins with wires are the solution of choice.

Supercritical-pressure, Once-through Cycle Light Water Cooled Reactor Concept

The purpose of the study is to develop new reactor concepts for the innovation of light water reactors (LWR) and fast reactors. Concept of the once-through coolant cycle, supercritical-pressure light water cooled reactor was developed. Major aspects of reactor design and safety were analysed by the computer codes which were developed by ourselves. It includes core design of thermal and fast reactors, plant system, safety criteria, accident and transient analysis, LOCA, PSA, plant control, start up and stability. High enthalpy rise as supercritical boiler was achieved by evaluating the cladding temperature directly during transients. Fundamental safety principle of the reactor is monitoring coolant flow rate instead of water level of LWR. The reactor system is compact and simple because of high specific enthalpy of supercritical water and the once-through cycle. The major components are similar to those of LWR and supercritical thermal plant. Their temperature are within the experiences in spite of the high outlet coolant temperature. The reactor is compatible with tight fuel lattice fast reactor because of the high head pumps and low coolant flow rate. The power rating of the fast reactor is higher than the that of thermal reactor because of the high power density.

Importance of Delayed Neutrons on the Coupled Neutronic-Thermohydraulic Stability of a Natural Circulation Heavy Water–Moderated Boiling Light Water–Cooled Reactor

The coupled neutronic-thermohydraulic stability characteristics of a natural circulation heavy water-moderated boiling light water-cooled reactor was investigated analytically considering the effects of prompt and delayed neutrons. For this purpose, the reactor considered is the Indian Advanced Heavy Water Reactor. The analytical model considers a point kinetics model for the neutron dynamics, a homogeneous two-phase flow model for the coolant thermal hydraulics, and a lumped heat transfer model for the fuel thermal dynamics. A higher mode of oscillation having a frequency much greater than the density-wave oscillation frequency was observed if prompt neutrons alone were considered. The occurrence of a higher mode of oscillation was found to be dependent on the concentration of delayed neutrons, the void reactivity coefficient, and the fuel time constant. The core inlet subcooling is found to have different effects on the decay ratio of the fundamental and higher modes of oscillations. The influences of void reactivity coefficient and fuel time constant on the fundamental and higher modes of oscillations were also found to be opposite in nature.

Startup Thermal Considerations for Supercritical-Pressure Light Water-Cooled Reactors

Supercritical-pressure light water-cooled reactors (SCRs) are innovative systems aimed at high efficiency and cost reduction. The once-through direct-cycle plant system is the leading system of fossil-fired power plants (FPPs). Estimates of the coolability and necessary sizes of the SCR startup systems, sequences, and required equipment for startup are investigated with reference to supercritical FPPs. There are two types of supercritical boilers. One is a constant pressure boiler, and the other is a variable pressure boiler.First, startup of the constant pressure boiler is examined. The reactor starts at a supercritical pressure. A startup bypass system consisting of a flash tank and pressure-reducing valves is required. Second, startup of the variable pressure boiler is investigated. The reactor starts at a subcritical pressure, and the pressure increases with the load. A steam-water separator and a drain tank are required for startup.The results of computer calculations show that with both constant pressure and variable pressure startup, the peak cladding temperature does not exceed the operating limit through startup, and both startup sequences are feasible. The sizes of the components required for the startup systems are assessed. To simplify the plant system and to reduce the component size, variable pressure startup with steam separators in the bypass line appears desirable.

Supercritical-pressure, Once-through Cycle Light Water Cooled Reactor Concept.

The purpose of the study is to develop new reactor concepts for the innovation of light water reactors (LWR) and fast reactors. Concept of the once-through coolant cycle, supercritical-pressure light water cooled reactor was developed. Major aspects of reactor design and safety were analysed by the computer codes which were developed by ourselves. It includes core design of thermal and fast reactors, plant system, safety criteria, accident and transient analysis, LOCA, PSA, plant control, start up and stability. High enthalpy rise as supercritical boiler was achieved by evaluating the cladding temperature directly during transients. Fundamental safety principle of the reactor is monitoring coolant flow rate instead of water level of LWR. The reactor system is compact and simple because of high specific enthalpy of supercritical water and the once-through cycle. The major components are similar to those of LWR and supercritical thermal plant. Their temperature are within the experiences in spite of the high outlet coolant temperature. The reactor is compatible with tight fuel lattice fast reactor because of the high head pumps and low coolant flow rate. The power rating of the fast reactor is higher than the that of thermal reactor because of the high power density.

Application of the pyrochemical process to recycle of actinides from LWR spent fuel

The Integral Fast Reactor (IFR) process includes recycle of all of the actinides contained in spent IFR fuel, and thus reduces waste disposal challenges. Pyrochemical processes also appears suitable for recovering and recycling actinides from spent fuels from light water cooled reactors (LWRs). Following surveys of many pyrochemical processes, two candidates have been selected for further development. A calcium/salt transport process operating at 800 to 850°C appears to recover the transuranic (TRU) elements efficiently. The oxide fuel is reduced to metal using calcium reductant. Salt extraction and transport are used to effect a suitable separation of transuranics from the excess uranium and the bulk of the fission products. The metallic TRU product is then used as feedstock for casting fast reactor fuel pins. An alternate, and currently preferred, process involves a lithium reductant operating at about 650°C. The electrorefining step used to separate the TRU product from the bulk excess uranium is precisely the one already developed for the IFR recycle. Basic laboratory work shows that either process would be feasible and engineering-scale experiments are being prepared.

Nonproliferation and safeguards aspects of the IFR

The National Academy of Sciences (NAS) has declared that the large and growing stocks of plutonium from weapons dismantlement in the U.S. and the former Soviet Union FSU are a “clear and present danger” to peace and security. Moreover, the opinion of some experts that plutonium of any isotopic blend is a proliferation threat (Mark, 1993) has been well publicized, heightening the concern that plutonium produced in the civilian fuel cycle is itself a proliferation threat. Assuring that separated plutonium, from dismantled warheads as well as from civilian power programs, is under effective control has (again) become a high priority of U.S. diplomacy. One pole of the debate on how to manage this material is to declare it to be a waste, and to search for some way to dispose of it safely, securely, and permanently. The other pole is to view it as an energy resource and to safeguard it against diversion, putting it into active use in the civilian power program. The ultimate choice cannot be separated from the long-term strategy for use of peaceful nuclear power. Continued use of a once-through fuel cycle will lead to an ever-increasing quantity of excess plutonium—requiring safeguarding. Alternatively, recycling the worl's stocks of plutonium in fast reactors, contrary to common misconception, will cap the world supply of plutonium and hold it in working inventories for generating power. Transition from the current-generation light water cooled reactors (LWRs) to a future fast-reactor-based nuclear energy supply under international safeguards would, henceforth, limit world plutonium inventories to the amount necessary and useful for power generation, with no further excess production. The IFR offers complete recycle of plutonium, and indeed, of all transuranics, with essentially no transuranics sent to waste, so the need for perpetual safeguards of Integral Fast Reactor (IFR) waste is eliminated. The pyro-recycle process is more proliferation resistant than the current plutonium-uranium extraction process (PUREX) because at every step of the IFR recycle process the materials meet the “spent-fuel standard”. The scale of IFR recycle equipment is compatible with colocation of power reactors and their recycle facility, eliminating off-site transportation and storage of plutonium-bearing materials. Self-protecting radiation levels are unavoidable at all steps of the IFR cycle, and the resulting limitation of access contributes to making covert diversion of material from an IFR very difficult to accomplish and easy to detect. Material diverted either covertly or overtly from the IFR fuel cycle would be difficult (relative to material available by other means, such as LWR spent fuel) to process into weapons feedstock.

Passive Pressure Tube Light Water Reactors

High-power-rating [{ge} 1,000 MW(electric)] passive pressure tube light water-cooled reactors are described that have the ability to reliably discharge decay heat to the ultimate heat sink, without the need to replenish primary coolant in loss-of-coolant accidents, while ensuring the integrity of the fuel and reusability of major reactor components. Evaluation shows that pressure tube reactors have the attractive potential to dissipate the decay heat from voided fuel elements of large-power-rating cores without exceeding safe temperature limits. Two basic versions of a pressure tube light water-cooled and -moderated reactor--the dry and wet calandria concepts--are proposed, and their advantages and limitations are discussed.

Conceptual Design Features for a Passive Light Water Cooled and Moderated Pressure Tube Reactor (PLPTR)

Key conceptual features are developed for a passive light water cooled and moderated pressure tube reactor having a total thermal output of 4029 MW (thermal) and net electric power of 1260 MW (electric) compatible with passive safety features that can ensure integrity of the fuel and reusability of the major reactor components without the requirement of delivery of emergency coolant makeup inventory. The concept of passive heat removal applied in this design combines heat storage, thermal conduction, radiation, natural circulation of light water in a moderator system under one atmosphere, and a natural draft air cooling system. Feasibility studies concerning normal heat transport, accident decay heat removal, materials compatibility, and reactor physics for criticality and feedback effects were carried out for the associated design features

Core thermohydraulic design with 20% LEU fuel for upgraded research reactor JRR-3.

This paper presents the outline of the core thermohydraulic design and analysis of the research reactor JRR-3, which is to be upgraded to a 20 MWt pool-type, light water-cooled reactor with 20% low enriched uranium (LEU) plate-type fuel. For the condition of normal operation, the upgraded JRR-3 core is planned to be cooled by two cooling modes of forced-convection at high power and natural-convection at low power. The major feature of core thermohydraulics is that at the forced-convection cooling mode the core flow is a downflow, under which fuel plates are exposed to a severer condition than an upflow in cases of operational transients and accidents. The core thermohydraulic design was, therefore, done for the condition of normal operation so that fuel plates may have enough safety margins both against the onset of nucleate boiling (ONR) not to allow the nucleate boiling anywhere in the core and against the departure from nucleate boiling (DNB). The safety margins against ONB and DNB were evaluated. The ...

新型転換炉原型炉,(II)

新型転換炉原型炉,(II)

新型転換炉原型炉,(I)

新型転換炉原型炉,(I)