Plutonium breeding of light water cooled fast reactors

Light water cooled fast reactor with new fuel assemblies (FA) has been studied for high breeding of fissile plutonium. It achieves fissile plutonium surviving ratio (FPSR) of 1.342 (discharge/loading), 1.013 end and beginning of equilibrium cycle (EOEC/BOEC), and compound system doubling time (CSDT) of 95.9 years at the average coolant density of pressurized water reactor (PWR). It is further improved for reduced moderation boiling water reactor (BWR) (RMWR) coolant density. Fissile plutonium surviving ratio reaches 1.397 (discharge/loading), 1.030 (EOEC/BOEC) and CSDT is 37 years. The present study has shown the possibility of breeding at the PWR coolant density and meeting the growth rate of energy demand of advanced countries at the RMWR and Super FR coolant density for the first time. The new FA consist of closely packed fuel rods. The integrity of welding of fuel rods at the top and bottom ends is maintained as the conventional fuel rods. The coolant to fuel volume fraction is reduced to 0.085, one-sixth of that of RMWR. The volume fraction remains unchanged with the diameter of the fuel rod. The thermal hydraulic design of the cores remains for the future study.     

Preliminary study on feasibility of large and small water cooled thorium breeder reactor in equilibrium states

Fuel breeding is one of the essential performances for a self-sustaining reactor system which can maintains the fuel sustainability while the reactor produces energy and consumes the fissile materials during operation. Thorium cycle shows some advantageous on higher breeding characteristics in thermal neutron spectrum region as shown in the Shippingport reactor and molten salt breeder reactor (MSBR) project. In the present study, the feasibility of large and small water cooled thorium breeder reactors is investigated under equilibrium conditions where the reactors are fueled by ²³³U–Th oxide and they adopts light water coolant as moderator. The key properties such as required enrichment, breeding capability, and initial fissile inventory are evaluated. The conversion ratio and fissile inventory ratio (FIR) are used for evaluating breeding performance. The results show the feasibility of breeding for small and large reactors. The breeding performance increases with increasing power output and lower power density. The small reactor may achieve the breeding condition when the fuel pellets' power density of about 22.5 W/cm³ and burnup of about 20 GWd/t.     

Study on the flow-pattern-transition instability in a natural circulation heavy water moderated boiling light water cooled reactor

A mathematical model has been developed to study the flow pattern transition instability which may occur in a boiling two-phase system. The model considers flow pattern transition criteria for vertical upward and horizontal flow in pipes to identify the flow pattern transition and flow pattern specific pressure drop models. It also considers the drift flux model to estimate the void fraction in the two-phase region. The model has been applied to predict the flow pattern transition instability in a natural circulation heavy water moderated boiling light water cooled reactor. It is found that the instability characteristics is similar to that of the Ledinegg-type instability. However, the number of multiple steady states for a given operating power can be much larger in the flow pattern transition instability as compared to that of the Ledinegg-type instability. Stability maps were plotted and compared for both the flow pattern transition instability and that of the Ledinegg-type instability. The influence of various geometric and operating parameters on this instability were investigated.     

Calculation of the moderator temperature coefficient of reactivity for water moderated reactors

An American nuclear standard has been recently issued, providing guidance and qualitative recommendations on how the moderator temperature coefficient (MTC) of reactivity should be determined through reactor calculations in water moderated (and cooled) reactor cores (ANS, 1997. Calculation and measurement of the moderator temperature coefficient of reactivity for water moderated power reactors. American Nuclear Society, American National Standard ANSI/ANS-19.11-1997). The present work provides quantitative information on areas of concern and effects addressed in this standard, namely, the effect of the core reflector region and the effect of the magnitude of the moderator temperature change. The calculations consist of a sequence of static criticality analyses, and are performed with a purposely developed two-dimensional reactor model based on two neutron energy groups. The numerical results indicate that the thickness of the reflector has a measurable effect on the accuracy of the MTC value only for a small core, introducing uncertainties of the order of 10%. Instead, the effect of temperature change is found to be negligible within the recommended range of 3 to 5°C.     

Determination of the external slug effect in multiplying systems with a heavy water moderator

The external slug effect has been determined in a heavy water multiplying system by measuring the thermal and epithermal neutron densities in a lattice cell. For the systems studied, the external slug effect and the neutron density distribution over the cell determined experimentally and by calculation are found to be in good agreement. Experiments have shown an appreciable sIug effect for epithermal neutrons.Translated from Atomnaya Énergiya, Vol. 14, No. 3, pp. 281–284, March, 1963     

Construction management of Indian pressurized heavy water reactors

Pandit Jawaharlal Nehru and Dr. Homi J. Bhabha, the visionary architects of Science and Technology of modern India foresaw the imperative need to establish a firm base for indigenous research and development in the field of nuclear electricity generation. The initial phase has primarily focused on the technology development in a systematic and structured manner, which has resulted in establishment of strong engineering, manufacturing and construction base.The nuclear power program started with the setting up of two units of boiling light water type reactors in 1969 for speedy establishment of nuclear technology, safety culture, and development of operation and maintenance manpower. The main aim at that stage was to demonstrate (to ourselves, and indeed to the rest of the world) that India, inspite of being a developing country, with limited industrial infrastructure and low capacity power grids, could successfully assimilate the high technology involved in the safe and economical operation of nuclear power reactors. The selection of a BWR was in contrast to the pressurized heavy water reactors (PHWR), which was identified as the flagship for the first stage of India's nuclear power program. The long-term program in three stages utilizes large reserves of thorium in the monazite sands of Kerala beaches in the third stage with first stage comprising of series of PHWR type plants with a base of 10,000 MW. India has at present 14 reactors in operation 12 of these being of PHWR type.The performance of operating units of 2720 MW has improved significantly with an overall capacity factor of about 90% in recent times.The construction work on eight reactor units with installed capacity of 3960 MW (two PHWRs of 540 MW each, four PHWRs of 220 MW each and two VVERs of 1000 MW each) is proceeding on a rapid pace with project schedules of less than 5 years from first pour of concrete. This is being achieved through advanced construction technology and management. Present efforts are focused on further reduction of gestation period. This is in contrast to construction period of 7–14 years in the earlier projects with labour intensive construction methods, learning period and indigenisation. The schedule and cost are interrelated and ultimately determine the viability and competitive edge of a project. With rich experience of over 30 years of operation and construction management it is well established that setting up of nuclear power projects in India in 4–5 years is quite feasible because of tremendous developments in construction technology; mechanization, parallel civil works and equipment erection, computerized project monitoring and accounting systems. By considering the best achieved times for the critical path activities of previous and ongoing projects, even a 4-year schedule is achievable. For nuclear power to be competitive it is essential that the gestation period is reduced and the capacity utilization enhanced. Both of these are the goals of the Indian nuclear power program. Presently the overnight cost per kW installed capacity is in the range of US$ 1100–1300 with levellised tariff of 5 c/kWh.     

Cold pressurisation in pressurised heavy water reactors during accidents

The paper addresses the issue of cold pressurisation during a Non-Loss of Coolant Accident (non-LOCA) and steam line break events. Cold pressurisation is usually avoided during reactor start up from cold and shut down to cold but it has to be avoided even during accidents so that consequential failures do not occur. A non-LOCA event, such as isolation of containment on a spurious signal and two main steam line breaks have been analysed. It is observed that the transient behaviour of the system is similar in these events. This paper discusses the issue of cold pressurisation in these events.     

Double-strata high burnup fuel performance in light water reactors

This study is aimed at the development of a fuel cycle concept for host countries with a lack of nuclear infrastructure. Two interrelated criteria, proliferation resistance and high-burnup, form the general framework of the fuel management scenario with the highest priority given to light water reactor technology and plutonium-free fresh fuel. Logically it implies the use of uranium oxide with enrichment close to 20%, whose effective utilization forms the main subject of the present paper. A sequence of two irradiation cycles for the same fuel pins in two different light water reactors is the key feature of the advocated approach. It is found that the synergism of PWR and pressure tube graphite reactor offers fuel burnup up to 140 GWd/tHM. Being as large as 8% in the final isotopic vector, the fraction of ²³⁸Pu serves as an inherent protective measure against plutonium proliferation.     

Conversion of high enriched uranium in thorium-232-based oxide fuel for light and heavy water reactors: MOX-T fuel

This paper discusses the possibility of using military high enriched uranium and plutonium in thorium oxide fuel for light and heavy water reactors (LWRs and HWRs). It is shown that such a fuel has several important advantages: (i) ²³⁹Pu and other long-living actinides are generated in quantities which are at least 100 times less than in conventional fuel; (ii) neutron emission is lower by a factor of more than 100; (iii) ²³³U is generated and burnt (the conversion factor for LWRs is 0.64–0.68 and for HWRs about 0.88); (iv) thorium is utilized and the total available amount of nuclear fuel is increased. The problem of non-proliferation of fissile material is also discussed and it is shown that the supervision of such fuel does not differ essentially from the supervision of low enriched uranium fuel with plutonium generation.     

Monitoring,diagnostics, and control system for large-capacity nuclear power facilities with water moderated and cooled reactors

The functional purpose and structure of the main information modules of an integrated data-acquisition and control system which performs monitoring, diagnostics, and control for nuclear power facilities with VVER reactors are described.     

Enhanced tritium production for fusion reactors via3He(n,p)3 H in the heavy water moderator of a CANDU reactor

There is an increasing requirement for tritium to supply the fuel needs of current experimental fusion devices and in the initial startup of future power generating reactors. Tritium is produced in heavy water reactors through deuterium activation, but the total production capacity of Canadian operated CANDUs will fall short of future demands, during the period before and for some time after self-sufficient reactors become available. Consequently, methods of enhancing tritium generating rates warrant investigation. Herein we provide the results of an inquiry into the feasibility of enhancing tritium production levels through the activation of helium-3 following its external addition to the heavy water moderator system of a hypothetical 500–600 MWe CANDU reactor. The approach adopted involves simulation of the temporal evolution of the tritium activities, originating from²H(n,)³H and³He(n, p)³H, as described by a simple first order kinetic model. The results suggest that the frequent addition of helium-3 to the moderator water will enhance tritium production inventories. The enhancement factor is highly dependent upon the rate at which helium-3 irretrievably escapes to the moderator cover gas. However, the direct activation of helium-3, contained in a closed loop such as the annulus gas system, for example, would be essentially complete within a few weeks without any significant loss.     

Review of thermal-hydraulic researches in severe accidents in light water reactors

The Tohoku Region Pacific Coast Earthquake and subsequent severe accident (SA) in Fukushima Daiichi Nuclear Power Station caused unprecedented disaster in Japan. Before this accident, considerable researches on SAs had been carried out in Japan. However, unfortunately, such researches could not prevent the accident due to the unexpected huge Tsunami. However, the researches on SAs become more and more important in order to make clear the causes of the accident in Fukushima and improve the safety of nuclear power plants in Japan. In view of this, review on researches on thermal hydraulics in SAs in light water reactors was carried out. Important thermal-hydraulic phenomena in SAs were identified. Research activities on each phenomenon were surveyed mainly based on the articles published in Journal of Nuclear Science and Technology of Atomic Energy Society of Japan.     

Western and Asian light water reactors at the millennium threshold

This paper presents the most advanced Western and Asian light water reactor (LWR) designs. The following pressurized water reactor (PWR) and boiling water reactor (BWR) designers are covered: Westinghouse ( ), Babcock and Wilcox (B&W—now part of Framatome), Combustion Engineering (CE—now ABB CE), Siemens (PWR), Framatome, Mitsubishi, General Electric (GE), Asea Brown Boveri (ABB), Siemens (BWR), Hitachi and Toshiba. The motivations that led to the design of the next generation of LWRs are discussed. The technical bases for evolutionary and innovative plants are summarized. Important safety features of some of the most complete (in operation, under construction or certified) evolutionary designs are described detail. Analogous implementations of systems into other advanced designs are given.     

Subchannel analysis of a fast reactor cooled by supercritical light water

The concept of a high temperature fast reactor cooled by supercritical water (SCFR-H) was developed for achieving high thermal efficiency and a compact reactor system. The core characteristics were obtained from single channel thermal-hydraulic analysis. Thus, it is necessary to carry out subchannel analysis to estimate the effect of local power peaking and cross flows. For this purpose, a subchannel analysis code is developed. It is verified by comparing the results with experimental data of High Conversion Pressurized Water Reactor (HCPWR). Sensitivities of the outlet coolant and cladding temperature to the subchannel flow area and local power peaking are high. One of the reasons is that the ratio of the coolant flow rate of SCFR-H to the power is smaller than that of LWR. Another reason is that, temperature of supercritical water is more sensitive to the enthalpy change above 450°C. The outlet coolant temperature distribution can be flattened by reducing the area of the peripheral subchannels and by enhancing the mixing between the subchannels.     

SBLOCA with boron dilution in pressurized water reactors. Impact on operation and safety

When the water level in the reactor pressure vessel (RPV) of a pressurized water reactor (PWR) is low enough and the core temperature is such that the coolant in that region boils, reflux-condensation conditions are established. Under such conditions, almost boron-free water is collected in a region of the primary system forming a non-borated slug. If subsequent natural circulation is established or a reactor coolant pump (RCP) is restarted, the slug could be transported to the core. This scenario configures an important part of the so-called boron issue. The Energy Systems Analysis Group at the Institute of Energy Technologies (INTE) of the Technical University of Catalonia (UPC) has studied the boron issue in three different stages. The steps were the following: participation in OECD-related projects, code improvement and investigation at nuclear power plant (NPP) scenarios. The third step is the main aim of this paper and consists of a continuation of the previous projects in the field of NPP analysis. The aim of this paper is to study SBLOCA transients with boron dilution in PWR. The chosen NPP was Ascó-2 which is a 3-loop-2940,6 MWth Westinghouse PWR. The paper contains some references to OECD/SETH and OECD/PKL experimental projects and analyses an established scenario including features of boron transport and sensitivity calculations for relevant parameters.     

Breeding and void reactivity analysis on heavy metal closed-cycle water cooled thorium reactor

Design parameters of heavy water (D₂O) cooled thorium breeder reactors for actinides closed-cycle cases have been investigated to find a design feasible area of breeding and negative void reactivity. Heavy metals (HMs) closed-cycle shows narrower feasible area compared with feasible area of ²³³U closed-cycle. In thorium fuel cycle, the breeding capability of the reactors becomes worse when all HMs are recycled. The result shows an opposite profile of breeding capability compared with uranium fuel cycle which obtains higher breeding capability when more HMs are recycled. Feasible design area which has a breeding and negative void reactivity can be estimated for higher burnup, even higher than 60 GW d/t for ²³³U closed-cycle; however, it is limited to 36 GW d/t for HM closed-cycle. Contribution of capture ²³⁵U is more significant to reduce breeding capability and contribution of ²³⁴U is also more effective to make the reactor more positive or less negative void coefficient for HM closed-cycle case in thorium fuel cycle system.     

Thermal-hydraulic transient analysis of light water reactors using moving boundaries

This paper examines the applicability of a mathematical dynamic model developed here for the simulation of the thermal-hydraulic transient analysis for light water reactors (LWRs). The thermal-hydraulic dynamic modeling of the fuel pin and adjacent coolant channel in LWRs is based on the moving boundary concept. The fuel pin model (FUELPIN) with moving boundaries is developed to accommodate the core thermal-hydraulic model, with detailed thermal conduction in fuel elements. Some results from transient calculations are examined for the first application of the thermal-hydraulic model and the fuel pin model with moving boundaries in a boiling water reactor (BWR). An accurate minimum departure from nucleate boiling ratio (MDNBR) and its axial MDNBR boundary versus time within the fuel channel are predicted during transients. Transient analysis using a known thermal-hydraulic code, COBRA and FUELPIN linked with a PWR systems analysis code show that the thermal margin gains more by a transient MDNBR approach than the traditional quasi-steady methodology for a pressurized water reactor (PWR). The studies of the overall nuclear reactor system show that moving boundary formulation provides an efficient and suitable tool for thermal transient analysis of LWRs.     

Comparison of thorium-based fuels with different fissile components in existing boiling water reactors

With the aim of investigating the technical feasibility of fuelling a conventional BWR (Boiling Water Reactor) with thorium-based fuel, computer simulations were carried out in a 2D infinite lattice model using CASMO-5. Four different fissile components were each homogenously combined with thorium to form mixed oxide pellets: Uranium enriched to 20% U-235 (LEU), plutonium recovered from spent LWR fuel (RGPu), pure U-233 and a mixture of RGPu and uranium recovered from spent thorium-based fuel. Based on these fuel types, four BWR nuclear fuel assembly designs were formed, using a conventional assembly geometry (GE14-N). The fissile content was chosen to give a total energy release equivalent to that of a UOX fuel bundle reaching a discharge burnup of about 55 MWd/kgHM. The radial distribution of fissile material was optimized to achieve low bundle internal radial power peaking. Reactor physical parameters were computed, and the results were compared to those of reference LEU and MOX bundle designs. It was concluded that a viable thorium-based BWR nuclear fuel assembly design, based on any of the fissile components, can be achieved. Neutronic parameters that are essential for reactor safety, like reactivity coefficients and control rod worths, are in most cases similar to those of LEU and MOX fuel. This is also true for the decay heat produced in irradiated fuel. However when Th is mixed with U-233, the void coefficient (calculated in 2D) can be positive under some conditions. It was concluded that it is very difficult to make savings of natural uranium by mixing LEU (20% U-235) homogenously with thorium and that mixing RGPu with thorium leads to more efficient consumption of Pu compared to MOX fuel.     

Risk assessment in setting inservice inspection priorities at light water reactors

As part of the Nondestructive Evaluation Reliability Program, sponsored by the U.S. Nuclear Regulatory Commission, Pacific Northwest Laboratory is developing a method that uses risk-based approaches to establish inservice inspection plans for nuclear power plant components. The method first uses probabilistic risk assessment (PRA) results and Failure Modes and Effects Analysis (FEMA) techniques to identify and prioritize the most risk-important systems and components for inspection. The acceptable level of risk from structural failure for important systems and components is then apportioned as a small fraction of the total PRA estimated risk for core damage. This process determines the target (acceptable) risk and failure probability values for individual components. The Surry Unit 1 Nuclear Power Station was selected for pilot applications of the method. The specific systems addressed are the reactor pressure vessel, the reactor coolant, the low-pressure injection, and the auxiliary feedwater. The results provide a risk-based ranking of components within these systems and relate the target risk to target failure probability values for individual components. These results will be used to guide the development of improved inspection plans for nuclear power plants.