Symbiosis system for transmutation on nitride fuel FBR and ADS

A conceptual scheme for mass flow of transmuting Plutonium (Pu), minor actinides (MA) and long-lived fission products (LLFP) is studied. In this feature, the existing light-water reactors (LWRs) cycle will be main stream for nuclear electric generation during a long-term period more than 50 years, and Pu will be utilized in mixed oxide fuel (MOX)-LWRs. In future, when Pu recycling system will be achived by introducing high-conversion LWRs (HCLWRs) and/or fast breeder reactors (FBRs), the accelerator driven transmutation system (ADS) transmutes Pu, MA and Iodine from Purex or Dry reprocessing. This is due to reduce burden for transmuting the excess or remained Pu, MA and LLFP by commercial reactor plants in Pu-recycling system. For this purpose, we introduce a concept of symbiosis system for transmutation based on nitride fuel FBR and ADS. The core design for lead-bismuth (Pb-Bi) cooled FBRs and ADS, Pb-Bi technologies, 15N enrichment and 14C toxicity are studied. And the mass flows for MA and Iodine are discussed based on an estimated scenario for nuclear electric plants introduction in future.     

A feasibility study on FP incineration for self-consistent nuclear energy system (SCNES)

A fast reactor core and fuel cycle concept is discussed for the future “Self-Consistent Nuclear Energy System (SCNES)” concept. The present study mainly discussed long-lived fission products (LLFPs) burning capability and recycle scheme in the framework of metal fuel fast reactor cycle, aiming at the goals for fuel breeding capability and confinement for TRU and radio-active FPs within the system. Combining neutron spectrum-shift for target sub-assemblies and isotope separation using tunable laser, LLFP burning capability is enhanced. This result indicates that major LLFPs can be treated in the additional recycle schemes to avoid LLFP accumulation along with energy production. In total, the proposed fuel cycle is a candidate for realizing SCNES concept.     

Nuclear Design Study on a Super High-burnup Fuel Assembly for Pressurized Water Reactors Using Transuranium

A conceptual design study was carried out on a super high-burnup mixed-oxide (MOX) fuel assembly (SHB FA) for pressurized water reactors (PWRs) using transuranium (TRU). This study aims to avoid the surplus plutonium (Pu) accumulation and to reduce the accumulation of long-lived radioactive minor actinides (MAS) by utilizing the currently existing PWRs under the condition that the Japanese program to develop fast breeder reactors (FBRs) is tend to delay. For this purpose, an SHB FA with discharged burnup of ?80 GWd/t was investigated by utilizing MAS positively as both burnable absorbers and fissile suppliers and loading high-content Pu. It is possible to load the SHB FAs in a current PWR together with UO₂ FAs and to use 2.5 times as much amount of Pu as that in a standard 1/3 MOX core. Moreover, it is found to be possible to reduce the total number of fresh FAs further from that of a high-burnup (55 GWd/t in maximum) UO₂ (4.9 wt%) core and also to reduce the accumulation of MAS in the nuclear fuel cycle significantly.     

Hydride fuel for LWRs—Project overview

This special issue of Nuclear Engineering and Design consists of a dozen papers that summarize the research accomplished in the DOE NERI Program sponsored project NERI 02-189 entitled “Use of Solid Hydride Fuel for Improved Long-Life LWR Core Designs”. The primary objective of this project was to assess the feasibility of improving the performance of pressurised water reactor (PWR) and boiling water reactor (BWR) cores by using solid hydride fuels instead of the commonly used oxide fuel. The primary measure of performance considered is the cost of electricity (COE). Additional performance measures considered are attainable power density, fuel bundle design simplicity, in particular for BWRs, safety, attainable discharge burnup, and plutonium (Pu) transmutation capability.Collaborating on this project were the University of California at Berkeley Nuclear Engineering Department (UCB), Massachusetts Institute of Technology Nuclear Science and Engineering Department (MIT), and Westinghouse Electric Company Science and Technology Department. Disciplines considered include neutronics, thermal hydraulics, fuel rod vibration and mechanical integrity, and economics.It was found that hydride fuel can safely operate in PWRs and BWRs having comparable or higher power density relative to typical oxide-fueled LWRs. A number of promising applications of hydride fuel in PWRs and BWRs were identified: (1) Recycling Pu in PWRs more effectively than is possible with oxide fuel by virtue of a number of unique features of hydride fuel-reduced inventory of ²³⁸U and increased inventory of hydrogen. As a result, the hydride-fueled core achieves nearly double the average discharge burnup and the fraction of the loaded Pu it fissions in one pass is double that of the MOX fuel. (2) Eliminating dedicated water moderator volumes in BWR cores, thus enabling significant increase of the cooled fuel rod surface area as well as the coolant flow cross-section area in a given fuel bundle volume while reducing the heterogeneity of BWR fuel bundles, thus achieving flatter pin-by-pin power distribution. The net result is an increase in the core power density and a reduction of the COE.A number of promising oxide-fueled PWR core designs were also found in this study: (1) The optimal oxide-fueled PWR core design features a smaller fuel rod diameter (D) of 6.5 mm and a larger pitch to rod diameter (P/D) ratio of 1.39 than that presently practiced by industry of 9.5 mm and 1.326. This optimal design can provide a 27% increase in the power density and a 19% reduction in the COE provided the PWR can be designed to have the coolant pressure drop across the core increased from the reference 0.20 MPa (29 psi) to 0.414 MPa (60 psi). Under the set of constraints assumed in this work, hydride fuel was found to offer comparable power density and economics as oxide fuel in PWR cores when using fuel assembly designs featuring square lattice and grid spacers. This is because pressure drop constraints prevented achieving sufficiently high power using hydride fuel with a relatively small P/D ratio of around 1.2 or less, where it offers the highest reactivity and a higher heavy metal (HM) loading. (2) Using wire-wrapped oxide fuel rods in hexagonal fuel assemblies, it is possible to design PWR cores to operate at ∼50% higher power density than the reference PWR design that uses grid spacers and a square lattice, provided 0.414 MPa coolant pressure drop across the core could be accommodated. Uprating existing PWRs to use such cores could result in up to 40% reduction in the COE. The optimal lattice geometry is D = 9.34 mm and P/D = 1.37. The most notable advantages of wire-wraps over grid spacers are their significantly lower pressure drop, higher critical heat flux, and improved vibration characteristics.The achievement of the highest power gains claimed in this study is possible as long as mechanical components like assembly hold-down devices (both in PWRs and in BWRs) and steam dryers (only in BWRs) are appropriately upgraded to accommodate the higher coolant pressure drop and flow velocities required for the high-performance LWR designs. The compatibility of hydride fuel with Zircaloy clad and with PWR and BWR coolants need yet be experimentally demonstrated. Additional recommendations are given for future studies that need to be undertaken before the commercial benefits from use of hydride fuel could be reliably quantified.     

Evaluation of recycle P&T treatment by using BWR to perform geologic disposal

Geologic disposal scenario combined with multiple-recycle P&T (partitioning and transmutation) treatment of MA (minor actinide) was thought to have an important potential merit to carry out the geologic disposal. P&T treatment was thought to have a role to avoid an uncertainty caused by geologic behavior, such as underground water migration rate, reductive and inorganic environment in a super-long time. Partitioning and grouping of LLRN (long-lived radionuclide), i.e. MA and LLFP (long-lived fission product), has an essential role to the transmutation treatment. B/T (burning and/or transmutation) treatment of MA, with R&P (reprocessing and partitioning) process, could be recycled and SLFP (short-lived fission product) was removed to immobilize in GSC (glass solidified canister). If R&P has a high performance, the multi-recycle system which combines B T-BWR (B/T boiling water reactor) and R&P can transmute of U&Pu, MA and LLFP etc. with small inventory. B/T fraction of MA is high in case of multi-recycle system, and then the remain of mass of MA or LLFP is recycled many times with low inventory. The B/T fraction of MA or LLFP could be high, if the flux is high, i.e. the time needed for high B/T fraction is relatively short in case of high flux BWR, or long in case of normal BWR. The optimum period for discharge of B/T fuel could be determined by the net difference between the reduced mass of MA burned and/or transmuted in the B/T fuel and the accumulated mass of MA in the normal fuel, and with the additional mass of MA accumulated by U&Pu, which was unrecovered in the reprocessing existed in B/T fuel.     

Recycle transmutation of MA and LLFP using BWR for sustaining geologic disposal

The actinides and fission products produced in nuclear fuels constitute an important part of the HLW. Therefore, methods for reducing the radiotoxicity of the MA and LLFP in HLW are presently under investigation. The purposes of this study are to evaluate the effectiveness of MA transmutation by taking advantage of neutron spectrum hardening due to void fraction along BWR axial direction; to understand the effectiveness of LLFP transmutation in BWR considering the large capture cross section of FP in thermal region; and to evaluate the macroscopic characteristics of longer residential period of LLFP target in the high burnup BWR core. Conceptual B/T BWR supposed in this study was reactor which the performance comparable to the current BWR. In MA transmutation case, the calculation was focused on varying the void fraction of 0 to 40% along the axial direction, which were directly associated to the lower and upper region of the BWR core. The performance of B/T BWR was evaluated in which four components of MA (²³⁷Np, ²⁴¹Am, ²⁴³Am, and ²⁴⁴Cm) with fixed fraction were blended with UO₂ in B/T fuel. While, for LLFP transmutation, the B/T BWR was assumed to have two homogeneous regions: {1} the region for UO₂ driver fuel (99% of fuel weight), and {2} the region for LLFP (⁹⁹Tc and ¹²⁹I) target capsules (1% of fuel weight), in which metallic Tc rods and iodine in the form of CeI₃ was contained in cylindrical target capsules. The evaluation functions are {1} fission-to-transmutation ratio, [F/T ratio]_MA, and {2} transmutation fraction, Tf_LLFP. Results show that the hardening neutron spectrum due to increase of void fraction in B/T BWR would result a higher [F/T ratio] of MA transmutation performance. Np and Am would be effectively loaded in the upper region of the core, while Cm could be loaded in any region of the core. At the EOC of equal or more than 50 GWd/Mg(HM), technetium has a higher transmutation fraction compared to iodine. To obtain higher LLFP transmutation fraction, the residential time in the LLFP targets in the core, should be kept for long time, for instance about 10 to 30 years. For that purpose, it was proposed that the number of B/T BWR system for LLFP treatment corresponds to the residential time of the LLFP target, i.e. 10 to 30 units.     

Research and Development of Minor Actinide-containing Fuel and Target in a Future Integrated Closed Cycle System

Research and development of minor actinide-containing fuels and targets, i.e., (Pu,Am)O₂–MgO, (Pu,Np)O₂–MgO, (U,Pu,Np)O₂, (U,Pu,Np)N and (Pu,Np,Zr)N, for use in a future integrated closed cycle system that includes fast reactor and accelerator driven sub-critical system is underway. The present statuses of fabrication test and property measurements are given. Design concept of the oxide target is described in detail together with a screening of the support material. A new apparatus for the measurement of mechanical properties at the elevated temperature is installed for use in evaluating the fuel-cladding mechanical interaction. Development histories with future prospects of two types of Np-containing fuels for the fast reactor are mentioned. Preliminary test results for a new nitride target for the accelerator driven sub-critical system are given. Finally, an irradiation test plan in the experimental fast reactor JOYO is briefly described.     

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.     

Neutronic Analysis for Transmutation of Minor Actinides and Long-Lived Fission Products in a Fusion-Driven Transmuter (FDT)

This study presents the transmutations of both the minor actinides (MAs: ²³⁷Np, ²⁴¹Am, ²⁴³Am and ²⁴⁴Cm) and the long-lived fission products (LLFPs: ⁹⁹Tc, ¹²⁹I and ¹³⁵Cs), discharged from high burn-up PWR-MOX spent fuel, in a fusion-driven transmuter (FDT) and the effects of the MA and LLFP volume fractions on their transmutations. The blanket configuration of the FDT is improved by analyzing various sample blanket design combinations with different radial thicknesses. Two different transmutation zones (TZ_MA and TZ_FP which contain the MA and LLFP nuclides, respectively) are located separately from each other. The volume fractions of the MA and the LLFP are raised from 10 to 20% stepped by 2% and from 10 to 80% stepped by 5%, respectively. The calculations are performed to estimate neutronic parameters and transmutation characteristics per D–T fusion neutron. The conversion ratios (CRs) for the whole of all MAs are about 65–70%. The transmutation rates of the LLFP nuclides increase linearly with the increase of volume fractions of the MA, and the ⁹⁹Tc nuclide among them has the highest transmutation rate. The variations of their transmutation rate per unit volume in the radial direction are quasi-concave parabolic.     

Transmutation of long-lived radioactive waste based on double-strata concept

A considerable attention is directed toward the reduction in the long-term potential hazard by partitioning and transmutation (P-T): separating long-lived nuclides from the waste stream and converting them into either shorter-lived or non-radioactive ones. The effects of higher Pu and minor actinide (MA) compositions on the transmutation rates have been studied for a typical mixed oxide (MOX)-fuel fast breeder reactor (FBR) core with 2600 MWt. The calculations showed that the transmutation rate for (Pu, MA) compositions from MOX -LWR becomes one half than that from UO₂-light water reactor (LWR). Furthermore, MA accumulation and transmutation based on Double-Strata Scenario have been investigated for introducing the accelerator driven transmutation system (ADS) with 800 MWt. It was shown that in the scenario of nuclear plant capacities for maximum 140 GWe, which consists of LWRs and FBRs, the introduction of ADS can play a significant role as “Transmuter” in the back-end of fuel cycle.     

Rapid Transmutation of High-Level Nuclear Wastes in a Catalyzed Fusion-Driven System

The aim of this study is to investigate the high-level waste (HLW) transmutation potential of fusion-driven transmuter (FDT) based on catalyzed D–D fusion plasma for various fuel fractions. The Minor actinide (MA) (²³⁷Np, ²⁴¹Am, ²⁴³Am and ²⁴⁴Cm) and long-lived fission product (LLFP) (⁹⁹Tc, ¹²⁹I and ¹³⁵Cs) nuclides discharged from high burn-up pressured water reactor-mixed oxide spent fuel are considered as the HLW. The volume fractions of the MA and LLFP are raised from 10 to 20% stepped by 2% and 10 to 80% stepped by 5%, respectively. The transmutation analyses have been performed for an operation period (OP) of up to 6 years by 75% plant factor (η) under a first-wall neutron load (P) of 5 MW/m² by using two different computer codes, the XSDRNPM/SCALE4.4a neutron transport code and the MCNP4B Monte Carlo code. The numerical results bring out that the considered FDT has a high neutronic performance for an effective and rapid transmutation of MA and LLFP as well as the energy generation along the OP.     

Studies on LLFP transmutation in a pressurized water reactor

A systematic study on the long-lived fission product (LLFP) transmutation in a pressurized water reactor (PWR) is performed, aiming at an optimal transmutation strategy for present nuclear energy development. The LLFPs selected in the analysis include ⁹⁹Tc and ¹²⁹I discharged from light water reactors (LWRs). The isotope ¹²⁷I is also considered to avoid the difficulties in isotopes separation. To minimize the negative impacts of LLFPs on the core performance and safety parameters, metallic technetium or MgI₂ target pins mixed with ZrH₂ are designed and investigated. Through the numerical analysis on equilibrium cycles, the transmuted amounts of ⁹⁹Tc and ¹²⁹I equal to the yields from 1.94 and 4.22 PWRs with a power of 1000 MWe, respectively. Numerical results indicate that both ⁹⁹Tc and ¹²⁹I can be transmuted conveniently in present PWRs in the form of target pins.     

Application of fluoride volatility method to the spent fuel reprocessing

ABSTRACT

An advanced reprocessing system has been developed to treat various SF (spent fuels): spent UO₂ and MOX (mixed oxide) fuels from LWR (light water reactor) and MOX fuel from FR (fast reactor). The system consists of SF fluorination to separate most U (uranium) as volatile UF₆, dissolution of solid residue containing Pu (plutonium), FP (fission products), MA (minor actinides) and partial U by nitric acid, and Pu+U separation from FP and MA by conventional solvent extraction. Gaseous UF₆ is purified by the thermal decomposition and the adsorption of volatile PuF₆ and adsorption of other impurities. This system is a hybrid process of fluoride volatility and solvent extraction and called FLUOREX. Fluorination of most U in the early stage of the reprocessing process is aimed at sharply reducing the amount of SF to be treated in the downstream aqueous steps and directly providing purified UF₆ for the enrichment process without conversion. The FLUOREX can flexibly adjust the Pu/U ratio, rapidly separate UF₆ and economically treat aqueous Pu+U. These features are especially suitable for the transition period fuel cycle from LWR to FR. This paper summarizes the feasibility confirmation results of FLUOREX.     

Time-dependent Neutronic Analysis for High Level Waste Transmutation in a Fusion-driven Transmuter

This study presents time-dependent transmutations of high-level waste (HLW) including minor actinides (MAs) and long-lived fission products (LLFPs) in the fusion-driven transmuter (FDT) that is optimized in terms of the neutronic performance per fusion neutron in our previous study. Its blanket has two different transmutation zones (MA transmutation zone, TZ_MA, and LLFP transmutation zone, TZ_FP), located separately from each other. High burn-up pressured water reactor (PWR)-mixed oxide (MOX) spent fuel is used as HLW. The time-dependent transmutation analyses have been performed for an operation period (OP) of up to 10 years by 75% plant factor (η) under a first-wall neutron load (P) of 5 MW/m². The effective half-lives of the MA and LLFP nuclides can be shortened significantly in the considered FDT while substantial electricity is produced in situ along the OP.     

钍基ADS快热耦合次临界核系统初步研究

为补偿由于次临界反应堆的燃耗所损失的反应性,降低次临界反应堆功率对加速器束流的依赖,考虑钍的转换,给出了采用钍基燃料,液态铅-铋合金单一回路冷却、石墨慢化的ADS快热单向耦合次临界堆芯设计方案。结果表明:本设计方案实现了堆芯功率展平、中子单向耦合,延长了换料周期,并消除了空腔的不利影响;堆芯寿期内的温度反应性反馈为负效应,安全性高;堆芯具有较高的能量放大能力;堆芯寿期内k_(eff)变化不超过1.05%;所需加速器最大束流强度为4.21mA;堆芯的MA嬗变支持比可达15个百万kW级的PWR,嬗变能力强。     

Burn-up characteristics of ADS system utilizing the fuel composition from MOX PWRs spent fuel

Burn-up characteristics of accelerator-driven system, ADS has been evaluated utilizing the fuel composition from MOX PWRs spent fuel. The system consists of a high intensity proton beam accelerator, spallation target, and sub-critical reactor core. The liquid lead–bismuth, Pb–Bi, as spallation target, was put in the center of the core region. The general approach was conducted throughout the nitride fuel that allows the utilities to choose the strategy for destroying or minimizing the most dangerous high level wastes in a fast neutron spectrum. The fuel introduced surrounding the target region was the same with the composition of MOX from 33 GWd/t PWRs spent-fuel with 5 year cooling and has been compared with the fuel composition from 45 and 60 GWd/t PWRs spent-fuel with the same cooling time. The basic characteristics of the system such as burn-up reactivity swing, power density, neutron fluxes distribution, and nuclides densities were obtained from the results of the neutronics and burn-up analyses using ATRAS computer code of the Japan Atomic Energy research Institute, JAERI.     

Evaluation on transmutation performance of minor actinides with high-flux BWR

The performance of high-flux BWR (HFBWR) for burning and/or transmutation (B/T) treatment of minor actinides (MA) and long-lived fission products (LLFP) was discussed herein for estimating an advanced waste disposal with partitioning and transmutation (P&T). The concept of high-flux B/T reactor was based on a current 33 GWt-BWR, to transmute the mass of long-lived transuranium (TRU) to short-lived fission products (SLFP). The nuclide selected for B/T treatment was MA (Np-237, Am-241, and Am-243) included in the discharged fuel of LWR. The performance of B/T treatment of MA was evaluated by a new function, i.e. [F/T ratio], defined by the ratio of the fission rate to the transmutation rate in the core, at an arbitrary burn-up, due to all MA nuclides. According to the results, HFBWR could burn and/or transmute MA nuclides with higher fission rate than BWR, but the fission rate did not increase proportionally to the flux increment, due to the higher rate of neutron adsorption. The higher B/T fraction of MA would result in the higher B/T capacity, and will reduce the units of HFBWR needed for the treatment of a constant mass of MA. In addition, HFBWR had a merit of higher mass transmutation compared to the reference BWR, under the same mass loading of MA.     

Pu recycling in a full Th-MOX PWR core. Part I: Steady state analysis

Current practice of Pu recycling in existing Light Water Reactors (LWRs) in the form of U-Pu mixed oxide fuel (MOX) is not efficient due to continuous Pu production from U-238. The use of Th-Pu mixed oxide (TOX) fuel will considerably improve Pu consumption rates because virtually no new Pu is generated from thorium. In this study, the feasibility of Pu recycling in a typical pressurized water reactor (PWR) fully loaded with TOX fuel is investigated.Detailed 3-dimensional 100% TOX and 100% MOX PWR core designs are developed. The full MOX core is considered for comparison purposes. The design stages included determination of Pu loading required to achieve 18-month fuel cycle assuming three-batch fuel management scheme, selection of poison materials, development of the core loading pattern, optimization of burnable poison loadings, evaluation of critical boron concentration requirements, estimation of reactivity coefficients, core kinetic parameters, and shutdown margin.The performance of the MOX and TOX cores under steady-state condition and during selected reactivity initiated accidents (RIAs) is compared with that of the actual uranium oxide (UOX) PWR core.Part I of this paper describes the full TOX and MOX PWR core designs and reports the results of steady state analysis. The TOX core requires a slightly higher initial Pu loading than the MOX core to achieve the target fuel cycle length. However, the TOX core exhibits superior Pu incineration capabilities.The significantly degraded worth of control materials in Pu cores is partially addressed by the use of enriched soluble boron and B₄C as a control rod absorbing material. Wet annular burnable absorber (WABA) rods are used to flatten radial power distribution. The temperature reactivity coefficients of the TOX core were found to be always negative. The TOX core has a slightly reduced, as compared to UOX core, but still sufficient shutdown margin.In the TOX core β_eff is smaller by about a factor of two in comparison to the UOX core and even lower than that of the MOX core. The combination of small β_eff and reduced control materials worth may potentially deteriorate the performance under RIA conditions and requires an additional examination. The behavior of the considered cores during the most limiting RIAs, such as rod ejection, main steam line break, and boron dilution, is further investigated and reported in Part II of the paper.     

Neutron Radiography Experiments at JRR-4, (II)

Experiments have been performed at JRR-4 (Japan Research Reactor) to investigate the capability of neutron radiographic techniques in applying to the nondestructive inspection of UO₂-PuO₂ mixed-oxide fuels. The object of the inspection was to detect “Pu particles” in the mixed-oxide fuels. In place of the actual fuel, two stages of TiO₂-EU₂O₃ mixed-oxide fuel dummy samples (the 1st stage for preliminary experiments and the 2nd stage for simulating the ATR fuel) were fabricated and radiographed with the direct exposure method using a Gd converter screen and high resolution films.

Neutron radiographs of the 1st stage dummy samples showed the excellent capability of the detection technique. Those of the 2nd stage dummy samples, however, revealed the detection limit of the technique, which showed that the present technique had not enough capability to satisfy the requirements in the inspection and that improvements of the detection technique especially on the contrast should be accomplished.     

乏燃料溶液嬗变堆焚烧锕系核素能力分析

针对焚烧锕系核素的目标,选择不同的乏燃料成分和堆芯功率,构造了7种乏燃料溶液嬗变堆( HSTR)堆芯模型,采用溶液堆堆芯燃料管理程序FMCHR计算了堆芯内Pu、Np及其他长寿命锕系核素的燃耗变化,分析了HSTR焚烧锕系核素的能力.结果表明:HSTR可以有效实现焚烧239pu的目标,同时嬗变可观数量的237Np;若要实现...