The concept and neutronic characteristics of a multipurpose fast reactor (MPFR)

The paper describes a concept and its neutronic characteristics of a long-life multipurpose nuclear reactor with self-sustained liquid metallic fuel, which is proposed to meet the requirements for the energy production in the future. The application of the liquid plutonium–uranium metallic alloys as a nuclear fuel demonstrates high potential to reach excellent conversion characteristics and high burn-up values with relatively small reactivity swings. Considerations about the capability of in-vessel separation of the main fission products to prolong the core lifetime without fuel reloading or shuffling are also discussed from the viewpoint of the influence on core burn-up characteristics.     

A concept of the multipurpose liquid metallic-fueled fast reactor system (MPFR)

Against fossil fuels, the nuclear energy is the only alternative energy source in the next century. Such energy source as the future nuclear power plant is expected to meet the following requirements. First, high temperature output for the multiple energy conversion capability as the electricity generation and the production of alternative fuels (hydrogen), which can be used widely in transportation systems. Second, the capability for siting close to the energy consumption area without onsite refueling. Third, the capability for nuclear fuel breeding and incineration of long-lived fission products, and fourth, the harmonization between active and passive safety features. This paper describes the basic concept of the Multipurpose liquid metallic-fueled Fast Reactor system (MPFR), which satisfies all mentioned requirements with introducing the U-Pu-x (x: Mn, Fe, Co) liquid metallic alloys for the fuel. We can obtain such characteristics as high operational temperature of the reactor (between 550 °C and 1200 °C) and elongation of the core operational lifetime by the inherent fission product separation in the liquid fuel by using these alloys. The enhanced self-controllability is achieved by the thermal expansion of liquid fuel; and the re-criticality phenomenon at the core compaction events can be eliminated by discharging of the liquid fuel from the core.     

Long life multipurpose small size fast reactor with liquid metallic-fueled core

A concept of a long life multipurpose nuclear reactor with self-sustained liquid metallic fuel is proposed to meet the requirements for the future energy production. The conceptual design is described and the core neutronic characteristics are obtained based on two-dimensional cylindrical diffusion reactor model. The influence of fission products separation in the liquid-fueled system on the core burnup capability is discussed. The burnup analysis shows a feasibility of the long life refueling-free core concept.     

Safety features of the gas-cooled fast breeder reactor

The safety features of the gas-cooled fast breeder reactor (GCFR) are described in the context of the 300-MW(e) demonstration plant design. They are of two general types, inherent and design-related. The inherent features are principally associated with the helium coolant and the nuclear coefficients. Design-related features influencing safety include shutdown systems, residual heat removal systems, method of core support, and the prestressed concrete reactor vessel (PCRV). This paper discusses the safety-related aspects of each of these. Recently completed residual heat removal system reliability studies are also discussed. The probability of residual heat removal system failure in the GCFR is found to be lower than that described for light water reactors. The safety characteristics of larger plants are examined, and increases in size are found to improve GCFR safety margins.     

中国实验快堆的安全特性

钠冷快堆因钠具有好的热物理特性而具有固有安全性,同时也因钠是活泼的碱金属,也难免会有钠的泄漏、钠火和钠水反应等工业事故.本文介绍了中国实验快堆利用钠冷快堆的固有安全性,装设了单靠自然循环和自然对流的事故余热导出系统等多项非能动安全系统及完善的能动安全系统,其安全性达到了第Ⅳ代先进核能系统的安全要求.对于大型快堆,因其保...     

Safety design of Pb–Bi-cooled direct contact boiling water fast reactor (PBWFR)

In Pb–Bi-cooled direct contact boiling water small fast reactor (PBWFR), steam is generated by direct contact of feedwater with primary Pb–Bi coolant above the core, and Pb–Bi coolant is circulated by steam lift pump in chimneys. Safety design has been developed to show safety features of PBWFR. Negative void reactivity is inserted even if whole of the core and upper plenum are voided hypothetically by steam intrusion from above. The control rod ejection due to coolant pressure is prevented using in-vessel type control rod driving mechanism. At coolant leak from reactor vessel and feedwater pipes, Pb–Bi coolant level in the reactor vessel required for decay heat removal is kept using closed guard vessel. Dual pipes for feedwater are employed to avoid leak of water. Although there is no concern of loss of flow accident due to primary pump trip, feedwater pump trip initiates loss of coolant flow (LOF). Injection of high pressure water slows down the flow coast down of feedwater at the LOF event. The unprotected loss of flow and heat sink (ATWS) has been evaluated, which shows that the fuel temperatures are kept lower than the safety limits.     

Safety analysis of small long life CANDLE fast reactor

Safety analysis for small long life fast CANDLE reactor was performed with ULOF (unprotected loss of flow), SDRW (unprotected shut down rods withdrawal), ULOHS (unprotected loss of heat sink) and LB (local blockage) accidents. The employed reactor system is based on the former steady state research. The core with 1.0 m radius and 2.0 m length produces 200 MW thermal power in steady state, using enriched N-15 natural uranium as fresh fuel and lead bismuth as coolant. The former 3 accidents were simulated without scram by neutronic-thermal hydraulic calculation coupled with stationary diffusion calculation. The LB accident was simulated by transient thermal hydraulic calculation only, because in this accident the neutronic factors basically do not change. The analysis results show that the proposed small CANDLE fast reactor can survive all the accidents without any active protection.     

Roughening characteristics and choices for the gas-cooled fast breeder reactor

The thermohydraulic performance of several types of rough surfaces proposed for use in the gas-cooled fast breeder reactor has been investigated experimentally at the Swiss Federal Institute for Reactor Research. Based on the tests, the most suitable roughness design has been defined. In addition to the thermohydraulic performance requirements, some other technological and operational criteria should be used for the final choice of roughness. There is not sufficient information on the different roughening methods to enable any decision to date, but when the new complex thermohydraulic performance criterion is considered, additional requirements become relatively more important.     

Some characteristics of a fast reactor with thorium shield

The regeneration factor of Pu²³⁹ in U²³³ was determined in the BR-1 experimentaI fast reactor with a Pu²³⁹ core and a Th²³² shield. The breeding characteristics of thorium, the utilization factor of fission neutrons (D) and the neutron multiplication factor (k) were also studied.Translated from Atomnaya Énergiya, Vol. 17, No. 4, pp. 294–299, October, 1964Deceased     

SSTAR: The US lead-cooled fast reactor (LFR)

It is widely recognized that the developing world is the next area for major energy demand growth, including demand for new and advanced nuclear energy systems. With limited existing industrial and grid infrastructures, there will be an important need for future nuclear energy systems that can provide small or moderate increments of electric power (10-700 MWe) on small or immature grids in developing nations. Most recently, the global nuclear energy partnership (GNEP) has identified, as one of its key objectives, the development and demonstration of concepts for small and medium-sized reactors (SMRs) that can be globally deployed while assuring a high level of proliferation resistance. Lead-cooled systems offer several key advantages in meeting these goals. The small lead-cooled fast reactor concept known as the small secure transportable autonomous reactor (SSTAR) has been under ongoing development as part of the US advanced nuclear energy systems programs. It is a system designed to provide energy security to developing nations while incorporating features to achieve nonproliferation goals, anticipating GNEP objectives. This paper presents the motivation for development of internationally deployable nuclear energy systems as well as a summary of one such system, SSTAR, which is the US Generation IV lead-cooled fast reactor system.     

Safety analysis of a supercritical-pressure water-cooled fast reactor under supercritical pressure

The supercritical-pressure water-cooled fast reactor (SWFR) is a fast spectrum supercritical water-cooled reactor (SCWR) studied by the University of Tokyo. The SWFR is designed as a two-pass core with an outlet temperature 500 °C. The SWFR has fuel channels cooled by downward flow, higher power density, and smaller coolant density reactivity feedback compared with Super LWR. This paper describes the safety analyses of abnormal events for the SWFR. SPRAT-F code is used for the safety analysis at supercritical pressure considering the downward flow cooled seed fuel channel. This code is based on a 1-D node junction model with point kinetics and decay heat calculations. Flow redistribution among parallel paths is calculated by pressure-loss balance and momentum conservation. The initiating events are selected from those of LWRs. For the safety analysis, nine abnormal transients and four accidents are selected with considering types of abnormality. By the numerical analyses, it was found that the loss of flow events can be mitigated by the “water source” effect of the downward flow blanket channels in the abnormal transients and accidents. All the abnormal events satisfy the criteria with margin.     

Transmutation characteristics of MA and LLFP in a fast reactor

Systematic studies were implemented to investigate the flexibility and attractive core concepts of MA and LLFP transmutation in fast reactors. The MA transmutation in the fast reactor core has no serious drawbacks in terms of core performance, provided that the homogeneous loading method can be employed with a small fraction of MA fuel (2˜5wt%). The recycling of MA in the fast reactor is feasible from neutronic and thermal-hydraulic points of view. For FP transmutation, the introduction of target subassemblies using duplex pellets — a moderator annulus surrounding a Tc-99 core — gives the maximum transmutation rate of Tc-99 in the radial shield region of the fast reactor. The fast reactor has an excellent potential for transmuting MA and LLFP effectively. The fast reactor will be able to play an important role for reduction of environmental burden in future energy system.     

Use of the calculated-expense method for choosing the characteristics of a fast reactor

The article describes how a calculated-expense formula can be used for analyzing the economics of the fuel cycle of a fast plutonium reactor. The proposed formula takes account of the down time in the fuel cycle for the fissionable material of the active zone and the blankets. It gives the results of the calculations, which are used for investigating how the fuel component of the calculated expenses depends on the power density and the flattening of the active zone of a reactor with an electrical power of 1000 MW.Translated from Atomnaya Énergiya, Vol. 21, No. 5, pp. 360–363, November, 1966.     

Commercialized fast reactor cycle systems and reactor core performance of the promising fast reactors

The Feasibility Study on Commercialized Fast Reactor (FR) Cycle Systems is under progress in order to propose prominent FR cycle systems that will respond to the diverse needs of society in the future. The design studies on various FR system concepts have been achieved and then the evaluations of potential to achieve the development targets have been also carried out. Crucial development issues have been found out for each FR system concept and their development plans for the key technologies are summarized as the roadmap. As a result, it has been confirmed that the sodium-cooled FR concept is highly suited to the development targets and R&D issues are related enhancing the economy with certain perspectives for realization. A flexible and robust development program for the FR cycle system will be proposed taking account of the characteristics for each FR concept until the end of the Phase II study.     

Core characteristics of fast reactor cycle with simple dry pyrochemical processing

Fast reactor core concept and core nuclear characteristics are studied for the application of the simple dry pyrochemical processing for fast reactor mixed oxide spent fuels, that is, the Compound Process Fuel Cycle, large FR core with half of loaded fuels are recycled by the simple dry pyrochemical processing. Results of the core nuclear analyses show that it is possible to recycle FR spent fuel once and to have 1.01 of breeding ratio without radial blanket region. The comparison is made among three kinds of recycle fuels, LWR UO₂ spent fuel, LWR MOX spent fuel, and FR spent fuel. The recycle fuels reach an equilibrium state after recycles regardless of their starting heavy metal compositions, and the recycled FR fuel has the lowest radio-activity and the same level of heat generation among the recycle fuels. Therefore, the compound process fuel cycle has flexibility to recycle both LWR spent fuel and FR spent fuel. The concept has a possibility of enhancement of nuclear non-proliferation and process simplification of fuel cycle.