一种铅-铋合金冷却快堆的高放废物积累量研究

目前商用压水堆积累了大量的长寿命高放废物,放射毒性强,衰变时间漫长,对环境和人类构成了长期威胁,作为6种第四代核能系统堆型中的一种,铅基冷却快堆在减少长寿命高放废物产生方面具有优势。基于此本文提出了一种热功率为300 MW的铅-铋合金冷却快堆设计。利用MCNP程序对反应堆堆芯进行建模并计算了堆芯在寿期初的主要物理参数,详细分析了燃耗过程中长寿命高放核素的积累量,并与一般压水堆长寿命高放核素的积累量进行了比较。结果表明,对主要关心的次锕系核素,铅-铋合金冷却快堆的产生量远小于压水堆的,而长寿命裂变产物的产生量与压水堆的相当。总体来说,铅-铋合金冷却快堆产生的长寿命高放废物总量小于压水堆的,可看出铅-铋合金冷却快堆在减少长寿命高放废物产生方面更具有竞争性。     

A concept of nitride fuel actinide recycle system based on pyrochemical reprocessing

A fuel cycle system coupled with nitride fuel fast reactors and a pyrochemical reprocessing has been investigated in order to establish an actinide transmutation recycle system with long-lived radioactive nuclides. Core performance of the nitride fuel fast reactor can provide design flexibility of excellent safety characteristics and a new concept of core composed with Na- and He- bonded fuel assemblies is proposed. The effect of ¹⁵N enrichment on nuclear characteristics and the evaluation of toxicity of ¹⁴C generated from ¹⁴N are appeared, and futhermore, excellent performance for the minor actinide (MA) transmutation is shown.

The study of the pyrochemical process shows that the actinides are reasonably separated from fission products in the nitride spent fuels, and that the high level wastes are of nearly actinide-free form.     

R&D activities based on fast reactor cycle technologies for transmutation of TRU and LLFP by JNC

Research and development(R&D) activities on partitioning and transmutation of trans-uranium nuclides (TRU) and LLFP and future R&D program in JNC were summarized. Feasibility design studies have been conducting to investigate the characteristics of a fast reactor core with TRU and LLFP transmutation. It was reconfirmed that the fast reactor has a strong potential for transmuting TRU and LLFP, effectively. R&D for establishing partitioning process of TRU apart from the high-level radioactive wastes have been carried out. By several counter-current runs of the TRUEX process using highly active raffinates, a process flow sheet capable of selective partitioning of actinides and fission products was newly developed. JNC has settled a new R&D program concerning partitioning and transmutation of long-lived radioactive waste based on recommendation of check & review for OMEGA program performed by the Ad Hoc Committee under the Atomic Energy Commission of Japan (AEC). The R&D program is composed of the design studies and development of element technologies (fabrication, irradiation) in the “Feasibility Studies” on commercialized fast reactor system and the basic studies with experiments (nuclear data, reactor physics, fuel property, etc.) to establish database and analytical tools for the TRU- and LLFP- containing fuel and core design.     

Study on MA and FP transmutation in fast reactors

Feasibility studies have been performed to develop an optimized fast reactor core for reducing long-term radiotoxicity of nuclear waste by minor actinide(MA) and long-lived fission product(FP) transmutation, taking into consideration fuel cycle technology. Systematic parameter survey calculations were implemented to investigate the basic characteristics of MA and FP transmutation in a fast reactor core. The hybrid MA-loading method, where Np nuclide is dispersed uniformly in the core and target subassemblies containing Am, Cm and rare earth nuclides are loaded into the blanket region, has the potential to achieve the maximum transmutation of MA with no special fuel design considerations. The introduction of target subassemblies using duplex pellets - a moderator annulus surrounding a ⁹⁹Tc core - has a great potential to transmute long-lived fission products in the radial blanket region of the fast reactor core.     

Burning transuranium isotopes in thermal and fast reactors

Energy production in nuclear power plants on the basis of fission processes lead inevitably to fission products and to the generation of new actinide isotopes. Most of these fission products are rather shortlived and decay within less than about 500 years to stable nuclides. However, a few of them, e.g. ⁹⁹Tc and ¹²⁹I, are longlived and may contribute to the radiotoxicity and hazard associated with an envisaged repository for their long-term disposal in a stable geologic formation, e.g. a salt dome. The majority of the generated actinide isotopes are fairly longlived, e.g. ²³⁹Pu with a halflife of more than 20 000 years. Therefore, their direct storage poses a heavy burden on the capacity and the possible environment impact of a repository. Furthermore, the energy content of these actinides could be deployed for producing additional nuclear fission energy after recovering them from unloaded irradiated fuel by suitable reprocessing techniques. Various possibilities exist for burning these actinides in different types of reactors, e.g. in light water reactors (LWRs), or LMFRs, adhering to available technology, or in actinide burners particularly designed for the purpose of their efficient incineration. The different options will be discussed in the paper. Transmutation of the manmade actinides and longlived fission products will require advanced technologies e.g. regarding reprocessing losses, remote fabrication techniques, and most probably, isotope separation processes. However, the almost complete elimination of these nuclides resulting from fission energy production in a continued recycling process may be the only feasible way to limit the effects of nuclear power generation to a tolerable and fair level for generations to come.     

加速器驱动次临界堆堆芯物理概念研究

分析了加速器驱动次临界堆芯的裂变核素增殖和平衡条件,主要长寿命放射性废物的积累,裂变产物毒性的影响及次临界堆的运行周期,输出功率和能量增益等主要性质,并对次临界热堆和次临界快堆的物理性质进行了比较。     

Possibility of long-term nuclear power development based on today’s technology

The possibility of long-term nuclear power development with a uranium fuel cycle based on ²³⁸U burnup and todays industrial technology is investigated. It is shown that such development is possible with fast reactors, including with sodium coolant. In this case, incomplete fuel reprocessing is admissable in a closed fuel cycle employing a pyroelectrochemical technology, which allows some fission products and actinides to be present in the fresh fuel prepared for reloading after reprocessing. These fission products and actinides can be burned in a reactor, thereby decreasing the quantity of radioactive wastes compared with the complete reprocessing with chemical separation of the fuel elements and decreasing the radiation load on the environment.Translated from Atomnaya Ènergiya, Vol. 97, No. 4, pp. 252–260, October, 2004.     

Was the Chazhma Accident a Chernobyl of the Far East?

The 1985 reactivity accident on a submarine in bukhta Chazhma was accompanied by a substantial emission of fission products and activational radionuclides whose total activity reached 5 MCi. Some specialists have compared this emission to the emission resulting from the 1986 accident in Chernobyl, neglecting the large difference in the radionuclide composition: short-lived products of prompt fission of uranium (with an admixture of activational ⁶⁰Co) in Chazhma and long-lived fission products accumulated over the run of the power reactor with an admixture of short-lived nuclides from the spontaneous excursion of the RBMK-1000 reactor in the Chernobyl nuclear power plant. It is shown that the emission of long-lived radioecologically significant radionuclides in Chazhma was approximately 0.79 Ci, while in the Chernobyl accident this emission was 90 MCi.A quantitative comparison is presented of the activity and radionuclide composition between the accidents in Chazhma and Chernobyl taking account of the fraction of long-lived radionuclides and neglecting the radioactive inert gases. These quantitative estimates are used to show that the Chazhma accident is not analogous to the 1986 accident in Chernobyl.     

Application of multiple γ-ray detection to long-lived radioactive nuclide determination in environmental samples

ABSTRACT

Gamma-gamma coincidence measurement utilized in γ-ray spectroscopy experiments is well known to be effective for the improvement of signal-to-noise ratio in a γ-ray spectrum. We study its applicability to the determination of long-lived radioactive nuclides in environmental samples. The γ-ray simulation code Geant 4.10.2 was used. A conventional and effective detector system comprising five Ge detectors was assumed. We took up 38 nuclides which need to be determined for the evaluation of fission product leakage at the nuclear accident in the Fukushima nuclear power plants in Japan. Among them 12 nuclides emit γ-rays and five nuclides of ⁶⁰Co, ⁹⁴Nb, ¹³⁴Cs, ¹⁵²Eu, and ¹⁵⁴Eu can be the objectives of the multiple γ-ray detection methods. The simulation results indicate that the signal-to-noise ratio can be improved by a factor between 9.84 and 283, and the detection limit by a factor between 2.71 and 8.53 relative to the singles measurement, implying that the method can be well applied to the determination of the long-lived radioactive nuclides and will provide a quick and non-destructive analysis method.     

The implication of high-flux fusion neutron source for the self-consistency in multi-component nuclear energy system

Among fission products (FP) discharged from a fission reactor, long-lived fission products are considered as of primary concern. Their transmutation has been of high priority to reduce the long-term consequences of nuclear energy generation. A self-consistent nuclear energy system (SCNES) in which we center fast breeder reactor may not have enough degree of excess neutron sources to transmute the fission products that potentially would poses environmental hazards in long-term period if they are buried in geologic disposal. Here we propose a so-called multi-component SCNES in which fission reactor systems can be combined with fusion reactor systems mainly for compensating the loss of enough capability for the transmutation. Amongst long-lived fission products, major concern has been paid for iodine and technetium and little attention was given to radioactive ⁹³Zr, although its hazard appears to be rather substantial. The importance of ⁹³Zr transmutation is emphasized and the transmutation capability was examined with fusion neutron sources by incorporating adequate moderation structures. As a result, we have demonstrated that the fusion neutron sources with high-flux blanket can be applied to transmute ⁹³Zr sufficiently and resolve the problem of its accumulation within the time period of several decades.     

Build-Up and Decay of Actinide Nuclides in Fuel Cycle of Nuclear Reactors

In order to assess the feasibility of utilizing plutonium in thermal reactors, build-up and decay of actinide nuclides have been studied for BWR, PWR, HWR, HTGR and LMFBR, which are uranium-oxide fueled or mixed-oxide fueled, and which produce electric power of 1,000MW. The following items were examined;

1. quantities of actinide nuclides build-up in the reactor

2. build-up and decay of activities of actinides in the spent fuel

3. build-up and decay of activities of actinides after reprocessing, and

4. variation of isotopie composition of plutonium with high burn-up.

It is concluded from the calculated results that precautions should be taken against high activities of resultant actinides if plutonium is utilized as a fissile material for thermal reactors. To make reprocessing and high-level waste management easy and practical, it is recommended that a thermal reactor should be fueled with uranium, the plutonium produced in a thermal reactor should be used in a fast reactor, and plutonium produced in the blanket of a fast reactor is more appropriate as fast reactor fuel than that from a thermal reactor.     

Release of radioactive materials from high active liquid waste in small-scale hot test for boiling accident in reprocessing plant

The release behavior of radioactive materials from high active liquid waste (HALW) has been experimentally investigated under boiling accident conditions. In the experiments using HALW obtained through laboratory-scale reprocessing, the release ratio was measured for fission product (FP) nuclides such as Ru, Tc-99, Cs, Sr, Nd, Y, Mo, Rh and actinides such as Cm-242 and Am-241. As a result, the release ratio was 0.20 for Ru and was around 1×10^?4 for the FP and actinide nuclides. Ru was released into the gas phase in the form of both mist and gas. For its released amount, weak dependency was found to its initial concentration in the test solution. The release ratio decreased with the increase in the initial concentration. For other FP nuclides and actinides as non-volatile, released into the gas phase in the form of mist, the released amount increased with the increase in the initial concentration. The release ratio of Ru and NOx concentration increased with the increase in the temperature of the test solutions. They were released together almost at the same temperature between 200 and 300 °C. Size distribution of particles like mist was measured. The data show that there was a difference between distributions at the temperatures below 150 °C and over 200 °C.     

Suitability of Small Scale Linear Systems for a Fission–Fusion Reactor, Breeder, and Waste Transmutation

To date the magnetic fusion effort has been almost entirely devoted to only one application, that being a multi gigawatt central station electric plant. Given the already well established fission based industry, the likelihood that fusion will have any impact on curbing the current carbon-based energy demand in the 21st century is slim. This paper advocates that the first and primary use of fusion neutrons should be as the driver for a sub-critical fission nuclear energy system—a fission–fusion hybrid reactor. This system can also be utilized to transmute long-lived radioactive wastes, and breed fissile nuclear fuel for several additional fission reactors. A small-scale fusion system based on a reciprocating fusion cycle employing the magneto-kinetic compression of the Field Reversed Configuration (FRC) is particularly well suited for this application. The characteristics of this fusion neutron driver and the potential for transmutation of long-lived nuclear wastes and breeding of fissile nuclear fuel in a blanket are presented.     

A metallic fuel cycle for Self-Consistent Nuclear Energy System (SCNES)

A fast reactor core and fuel cycle concept has been discussed for Self-Consistent Nuclear Energy System (SCNES) concept. This paper discussed loading material candidates for long-lived fission products (LLFPs) and removal of stable nuclides from radioactive nuclides with isotope separation using tunable laser. Some of LLFPs were possible to be loaded in metal of the generated form. The potential for LLFP-confinement in the reactor system is discussed along with a metallic fuel cycle concept. The proposed fuel cycle scheme is a successful candidate for SCNES concept.     

Calculation of Critical Concentrations of Actinides in an Infinite Medium of Silicon Dioxide

The critical concentrations of actinides in metal-silicon-dioxide (SiO₂) and in metal-water (H₂O) mixtures were calculated for 26 actinides including ^233,235U, ^239;241Pu, ^242mAm, ^243;245;247Cm, and ^249;251Cf. The calculations were performed using the Monte Carlo neutron transport calculation code MCNP5 combined with the evaluated nuclear data library JENDL3.3. The results showed that the critical concentration of actinide in metal-SiO₂ mixtures was about 1/5 of that in metal-H₂O mixtures for all the fissile nuclides investigated. The k_∞'s of metal-SiO₂ and metal-H₂O at one-half of the respective critical concentration of actinide, which was assumed as the subcritical concentration limit, were found to be less than 0.8 for all the actinides considered. By applying the sum-of-fractions rule to the concentrations of six nuclides in metal-SiO₂ mixtures, the subcriticality of high-level radioactive wastes was confirmed for a reported sample. The effects of different nuclear data libraries on the results of critical concentrations were found to be large for ²⁴²Cm, ²⁴⁷Cm, and ²⁵⁰Cf by comparison with the results calculated with another evaluated nuclear data library, ENDF/B-VI.     

Sensitivity of charged particle activation analysis for long-lived radioactive nuclide determination

ABSTRACT

Charged Particle Activation Analysis (CPAA) utilizing an 8-MeV proton beam has been studied for determination of 35 long-lived radioactive nuclides. We accumulated the reaction cross section and nuclear decay data by referring to nuclear database supplied by National Nuclear Data Center in Brookhaven National Laboratory. We also calculated the reaction cross sections by using statistical model code ALICE. By using the nuclear data, we have derived determination sensitivity of the radioactive nuclides relative to unit weight and specific radioactivity. The result indicates that several hardly measurable nuclides with long half-lives such as ¹³⁵Cs, ²⁴⁴Pu, ¹²⁹I, ¹²⁶Sn, ⁹³Mo, ¹⁰⁷Pd, ²³⁶U, ²⁴⁸Cm, and ²³⁷Np have high sensitivity. It may be concluded that CPAA can be applied to determination of several long-lived nuclei and will provide a quick and non-destructive analysis method.     

The Cost of Transmutation of Fission Products in Nuclear Reactors

The operating regime of a VVÉR reactor in which the most important long-lived fission products ⁹⁹Tc and ¹²⁹I are transmuted is investigated. Estimates are presented for the decrease in the fuel burnup and decrease in the run time as a result of transmutation. Two methods for inserting the nuclides to be transmuted are examined – by adding to the nuclear fuel or the coolant. It is established that ⁹⁹Tc and ¹²⁹I transmutation with the rate of accumulation in a reactor decreases burnup by 5.1 GW·days/metric ton, i.e., by 12.7% of the standard burnup. This corresponds to electricity underproduction 110 GW·days per run or 37 GW·days per year of operation. This result is independent of the method used to insert the nuclides to be transmuted. These energy losses are the price to be paid for transmuting nuclides without removing them during reactor operation.     

Fission Product Model for BWR Lattice Calculation Code

A benchmark calculation of full fission product was performed for thermal reactor application using an isotope transmutation code DCHAIN based on 185 nuclides with revised nuclear data library. The fission product model for BWR lattice calculation was studied and tested with the benchmark results, and a model containing 45 explicit nuclides and one pseudo nuclide was selected as a reasonably best model to predict the burn up reactivity with high precision for practically all types of fuel and reactor operating conditions. The evaluated thermal cross section and resonance integral for the pseudo nuclide are σ_2,200 = 2.6b and.RI = 10.6b, combined with the pseudo fission yield values of 1.3898, 1.3233, 1.3675 and 1.2773 for fissions from ²³⁵U, ²³⁸U, ²³⁹Pu and ²⁴¹Pu, respectively. The present results are believed as equally applicable to PWR lattice calculation.