Measurements of Neutron Capture Cross Section of 237Np for Fast Neutrons

The neutron capture cross section of ²³⁷Np has been measured for fast neutrons supplied at the center of the core in the Yayoi reactor. The activation method was used for the measurement, in which the amount of the product ²³⁸Np was determined by γ-ray spectroscopy using a Ge detector. The neutron flux at the center of the core calculated by the Monte Carlo simulation code MCNP was renormalized by using the activity of a gold activation foil irradiated simultaneously. The new convention is proposed in this paper to make possible a definite comparison of the integral measurement by the activation method using fast reactor neutrons with differential measurements using accelerator-based neutrons. “Representative neutron energy” is defined in the convention at which the cross section deduced by the activation measurement has a high sensitivity. The capture cross section of ²³⁷Np corresponding to the representative neutron energy was deduced as 0:80 ± 0:04b at 214 ± 9 keV from the measured reaction rate and the energy dependence of the cross section in the nuclear data library ENDF/B-VII.0. The deduced cross section of ²³⁷Np at the representative neutron energy agrees with the evaluated data of ENDF/B-VII.0, but is 15% higher than that of JENDL-3.3 and 13% higher than that of JENDL/AC-2008.     

The potential use of Am as proliferation resistant burnable poison in PWRs

This paper discusses the use of ²⁴¹Am as proliferation resistant burnable poison for light water reactors. Homogeneous addition of small (as little as 0.12%) amounts of ²⁴¹Am to the conventional light water reactor fuel results in significant increase in ²³⁸Pu/Pu ratio in the discharged fuel improving its proliferation resistance. Moreover, ²⁴¹Am, admixed to the fuel, acts as burnable absorber allowing for substantial reduction in conventional reactivity control means without a notable fuel cycle length penalty. This is possible due to favorable characteristics of ²⁴¹Am transmutation chain. The fuel cycle length penalty of introducing ²⁴¹Am into the core is evaluated and discussed, as well as the impact of He production in the fuel pins and degradation of reactivity feedback coefficients. Proliferation resistance and reactivity control features related to the use of ²⁴¹Am are compared to those of using ²³⁷Np, which has also been suggested as an additive to the conventional fuel in order to improve its proliferation resistance. It was found that ²⁴¹Am admixture is more favorable than ²³⁷Np admixture because of the smaller fuel cycle length penalty and higher burnable poison savings. Addition of either ²³⁷Np or ²⁴¹Am would provide substantial but not ultimate protection from misuse of Pu originating in the spent fuel from the commercial power reactors. Therefore, the benefits from application of the concept would have to be carefully evaluated against the additional costs and proliferation risks associated with manufacturing of ²³⁷Np or ²⁴¹Am doped fuel. Although this work concerns specifically with PWRs, the conclusions could also be applied to BWRs and, to some extent, to other thermal spectrum reactor types.     

Advanced U–Np–Pu fuel to achieve long-life core in heavy water reactor

The objective of this paper is to look at the possibility of approaching the long-life core comparable with reactor life-time. The main issues are centered on U–Np–Pu fuel in a tight lattice design with heavy water as a coolant. It is found that in a hard neutron spectrum thus obtained, a large fraction of ²³⁸Pu produced by neutron capture in ²³⁷Np not only protects plutonium against uncontrolled proliferation, but substantially contributes in keeping criticality due to improved fissile properties (its capture-to-fission ratio drops below unit). Equilibrium fuel composition demonstrates excellent conversion properties that yield the burn-up value as high as 200 GWd/t at extremely small reactivity swings.     

Separation of Plutonium-238 from Fission Products by Solvent Extraction Using HDEHP

A procedure for separating ²³⁸Pu from a Np sample irradiated with neutrons is described. Rapid separation of Pu by HDEHP solvent extraction was attempted, and without adjusting its valency states in the dissolver solution of the sample. Both Pu(IV) and Pu(VI) were extracted along with Np from the HNO₃ solutions of various concentrations. The Pu and Np extracted in the organic solution were back-extracted with oxalic acid solutions. The decontamination factors of the crude products were of the order of 10² for gross γ-activity. The Pu in the products was separated from Np by means of ion exchange resin columns. Approximately 0.5 mg of ²³⁸Pu was obtained with an efficiency exceeding 95%.     

Development of an Improved Ceramic Tritium Breeder

The paper presents the advanced concept of a long-life small light water reactor in which the fuel irradiation time is comparable with reactor life-time. The equilibrium analysis reveals that the U-Np-Pu fuel with unique neutronic properties allows to keep sufficient criticality up to burnup value about 140GWd/tHM. The fuel recycle does not lead to additional Pu accumulation. Both Pu and Np are well protected against un-controlled proliferation by a large fraction of ²³⁸Pu in their mixture. To improve the reactor safety, the wider fuel pin lattice was applied. The radiation damage of structural materials is within the stainless steel limitation.     

Irradiation analysis of U, Am samples irradiated in experimental fast reactor “Joyo” for protected plutonium production in fast breeder reactor blanket

Americium is a key element to design the FBR based nuclear fuel cycle, because of its long-term high radiological toxicity as well as a resource of even-mass-number plutonium by its transmutation in reactors, which contributes the enhancement of proliferation resistance. The present paper summarizes analysis of the individual Am and U samples irradiation in Joyo to re-evaluate the results of Pu isotopes in the measure of proliferation resistance, and to combine the results for the prediction of DU-Am irradiation especially in the production of Pu isotopes. By the prediction of DU-Am oxide fuel in fast reactor environment with detail fuel irradiation analysis, it was confirmed that neutron moderation and fuel size affects the produced Pu isotope and its vector due to the very high sensitivity of ²³⁸U resonance capture reaction, the larger diameter fuel is more preferable in the case of moderated neutron spectrum environment for denaturing Pu in fast reactor blanket. Finally proliferation resistance of all the Pu produced in U, Am sample irradiation and DU-Am fuel irradiation prediction were evaluated based on decay heat and spontaneous fission neutron rate, and it was confirmed ²⁴¹Am produces un-attractive Pu to abuse from the beginning to the end of irradiation, and more than 2% of ²⁴¹Am doping is required to enhance the proliferation resistance of Pu to MOX grade and Kessler’s proposal in moderated neutron spectrum environment in fast reactor.     

Analyses with latest major nuclear data libraries of the fission rate ratios for several TRU nuclides in the FCA-IX experiments

In the 1980s, a series of integral experiments was conducted in FCA-IX assemblies with systematically changed neutron spectra covering from the intermediate to fast ones. The experiments provide systematic data of central fission rates for TRU nuclides containing minor actinides, ²³⁷Np, ²³⁸Pu, ²³⁹Pu, ²⁴²Pu, ²⁴¹Am, ²⁴³Am, and ²⁴⁴Cm. Regarding the fission rate ratios relative to ²³⁹Pu, benchmark models had been recently developed for validation of nuclear data for the TRU's fission cross sections. In this paper, the latest major nuclear data libraries, JENDL-4.0, ENDF/B-VII.1, and JEFF-3.2, are compared on the benchmark models. For the libraries, the analyses by a Monte Carlo calculation code show obvious overestimations particularly for the fission rate ratios of ²⁴⁴Cm to ²³⁹Pu. Additionally, a large discrepancy about by 20% between the libraries is revealed for the fission rate ratio of ²³⁸Pu to ²³⁹Pu measured in the intermediate neutron spectrum. The causes of discrepancies are furthermore clarified by sensitivity analyses.     

~(238)Pu低能光子源衰变γ能谱识别

利用同轴P型高纯锗探测器,对X荧光分析的~(238)Pu低能光子源进行γ能谱分析,并对~(233)Pa、~(224)Ra、~(212)Pb、~(212)Bi及~(208)Tl的特征γ射线进行分析,确定上述核素的来源。其中,~(233)Pa是生产~(238)Pu的原料237 Np的衰变产物,~(224)Ra、~(212)Pb、~(212)Bi及~(208)Tl均为生产~(238)Pu的副产物~(236)Pu的衰变子核。能量为350、440、844、1 014、1 130、1 266、1 368、1 454keV的γ射线是α粒子轰击源封装材料引起原子核库伦激发或γ射线照射周边环境引起核激发产生。进行效率刻度后,使用γ能谱法计算各放射性核素的活度,并根据放射性平衡计算各放射性核素的质量。通过对~(238)Pu源γ能谱的分析,建立计算放射性同位素活度与质量的方法。     

Inherent Protection of Plutonium by Doping Minor Actinide in Thermal Neutron Spectra

The present study focuses on the exploration of the effect of minor actinide (MA) addition into uranium oxide fuels of different enrichment (5% ²³⁵U and 20% ²³⁵U) as ways of increasing fraction of even-mass-number plutonium isotopes. Among plutonium isotopes, ²³⁸Pu, ²⁴⁰Pu and ²⁴²Pu have the characteristics of relatively high decay heat and spontaneous fission neutron rate that can improve proliferation-resistant properties of a plutonium composition. Two doping options were proposed, i.e. doping of all MA elements (Np, Am and Cm) and doping of only Np to observe their effect on plutonium proliferation-resistant properties. Pressurized water reactor geometry has been chosen for fuels irradiation environment where irradiation has been extended beyond critical to explore the subcritical system potential. Results indicate that a large amount of MA doping within subcritical operation highly improves the proliferation-resistant properties of the plutonium with high total plutonium production. Doping of 1% MA or Np into 5% ²³⁵U enriched uranium fuel appears possible for critical operation of the current commercial light water reactor with reasonable improvement in the plutonium proliferation-resistant properties.     

Enhancing BWR proliferation resistance fuel with minor actinides

To reduce spent fuel for storage and enhance the proliferation resistance for the intermediate-term, there are two major approaches (a) increase the discharged spent fuel burnup in the advanced light water reactor- LWR (Gen-III Plus), which not only can reduce the spent fuel for storage, but also increase the ²³⁸Pu isotopes ratio to enhance the proliferation resistance, and (b) use of transuranic nuclides (²³⁷Np and ²⁴¹Am) in the high burnup fuel, which can drastically increase the proliferation resistance isotope ratio of ²³⁸Pu/Pu. For future advanced nuclear systems, minor actinides (MA) are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. As a result, MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. In the study, a typical boiling water reactor (BWR) fuel unit lattice cell model with UO₂ fuel pins will be used to investigate the effectiveness of minor actinide reduction approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance in the intermediate-term goal for future nuclear energy systems. To account for the water coolant density variation from the bottom (0.76 g/cm³) to the top (0.35 g/cm³) of the core, the axial coolant channel and fuel pin were divided to 24 nodes. The MA transmutation characteristics at different elevations were compared and their impact on neutronics criticality discussed. The concept of MARA, which involves the use of transuranic nuclides (²³⁷Np and/or ²⁴¹Am), significantly increases the ²³⁸Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in atoms for peace and the intermediate-term of nuclear energy reconnaissance.     

An experimental investigation of accumulation and transmutation behavior of americium in the MOX fuel irradiated in a fast reactor

Americium isotopes generated in the MOX fuel irradiated in the experimental fast reactor JOYO were analyzed by applying a sophisticated radiochemical technique. Americium was isolated from the irradiated MOX fuel by a combined method of anion-exchange chromatography and oxidation of Am. The isotopic ratios of americium and its content were determined by thermal ionization mass spectroscopy and α-spectrometry, respectively. The americium isotopic ratio was similar for all the specimens, but was significantly different from that of PWR-MOX. On the basis of present analytical results, the accumulation and transmutation behavior of americium nuclides in a fast reactor is discussed from the viewpoints of neutron spectrum dependence and the isomeric ratio of the ²⁴¹Am capture reaction. The estimated isomeric ratio is about 87%, which is close to the latest evaluated value. A rapid estimation method of Am content by using the ²⁴⁰Pu to ²³⁹Pu ratio was adopted and proved to be valid for the spent fuel irradiated in the fast reactor.     

Measurements of thermal-neutron capture cross-section and resonance integral of neptunium-237

The thermal-neutron capture cross-section (σ₀) and resonance integral (I₀) were measured for the ²³⁷Np(n,γ) ²³⁸Np reaction by an activation method. A method with a Gadolinium filter, which is similar to the Cadmium difference method, was used to measure the σ₀ with paying attention to the first resonance at 0.489 eV of ²³⁷Np, and a value of 0.133 eV was taken as a cut-off energy. Neptunium-237 samples were irradiated at the pneumatic tube of the Kyoto University Research Reactor in Institute for Integral Radiation and Nuclear Science, Kyoto University. Wires of Co/Al and Au/Al alloys were used as monitors to determine thermal-neutron fluxes and epi-thermal Westcott’s indices at an irradiation position. A γ-ray spectroscopy was used to measure activities of ²³⁷Np, ²³⁸Np and neutron monitors. On the basis of Westcott’s convention, the σ₀ and I₀ values were derived as 186.9 ± 6.2 barn, and 1009 ± 90 barn, respectively.     

Fission-product yields from neutron-induced fission

Exsting experimental thermal, fast, and 14-MeV neutron-induced fission-product cumulative and independent yieds have been compiled, corrected to common reference values, and listed in tabular form for the following fissile nuclides:Thermal-neutron fission: cumulative yields for ²²⁷Th, ²²⁹Th, ²³³U, ²³⁵U, ²³⁹Pu, ²⁴¹Pu, ²⁴¹Am, ²⁴²Am, ²⁴⁵Cm, ²⁴⁹Cf, ²⁵¹Cf, ²⁵⁴Es, and ²⁵⁵Fm; independent yieds for ²³³U, ²³⁵U, ²³⁷Np, ²³⁸U, and ²³⁹Pu.Fast-neutron fission: cumulativ yields for ²²⁷Ac, ²³¹Pa, ²³²Th, ²³³U, ²³⁵U, ²³⁷Np, ²³⁸U, and ²³⁹Pu; independent yields for ²³⁵U and ²³⁸U.14-MeV-neutron fission: cumulative yields for ²³¹Pa, ²³²Th, ²³³U, ²³⁵U, ²³⁷Np, ²³⁸U, and ²³⁹Pu; independent yields for ²³²Th, ²³³U, ²³⁵U, ²³⁸U, and ²³⁹Pu.11-MeV-neutron fission: cumulative yields for ²³²Th.3-MeV-neutron fission: cumulative yields for ²³¹Pa, ²³²Th, and ²³⁸U.1.1-MeV-neutron fission: cumulative yields for ²³⁷Np.From these experimental values the unknown independent yields are deduced empirically for thermal-neutron fission of ²³³U, ²³⁵U, ²³⁹Pu, and ²⁴¹Pu; the fast fission of ²³²Th, ²³³U, ²³⁵U, ²³⁸U, ²³⁹Pu, ²⁴⁰Pu, and ²⁴¹Pu (the chain yields for ²⁴⁰Pu and ²⁴¹Pu used at this energy being predictions); and the 14-MeV-neutron fission of ²³²Th, ²³³U, ²³⁵U, and ²³⁸U.Finally, by the fitting of the preceding information to condition equations derived from the conservation laws, adjusted sets of chain and independent yields are calculated for thermal fission of ²³³U, ²³⁵U, ²³⁹Pu, and ²⁴¹Pu; fast fission of ²³²Th, ²³³U, ²³⁵U, ²³⁸U, ²³⁹Pu, and ²⁴¹Pu; and 14-MeV fission of ²³²Th, ²³³U, ²³⁵U, and ²³⁸U. The literature search is probably complete to the end of 1975; some 1976 results are included.This paper replaces and makes obsolete the following UKAEA reports: AERE-R7209, AERE-R7394, AERE-R7680, and AERE-R8152.     

Measurement of Thermal Neutron Capture Cross Section and Resonance Integral of the 237Np(n, γ)238Np Reaction

The thermal neutron capture cross section (σ₀) and the resonance integral (I ₀)of ²³⁷Np have been measured by an activation method to supply basic data for the study of transmutation of nuclear waste. The neutron irradiation of ²³⁷Np samples have been done at the Research Reactor Institute, Kyoto University (KUR). Samples of ²³⁷Np were irradiated between two Cd sheets or without a Cd sheet. Since ²³⁷Np has a strong resonance at the energy of 0.49 eV, the Cd cutoff energy was adjusted at 0.358 eV (thickness of the Cd sheets: 0.125 mm). A high purity Ge detector was employed for activity measurement. The reaction rate to produce ²³⁸Np from ²³⁷Np was analyzed by the Westcott's convention. Results obtained were 141.7±5.4 barns for σ₀ and 862±51 barns for I ₀ above 0.358 eV of ²⁷³Np. By setting the Cd cut-off energy at 0.358 eV considering the resonance at 0.49 eV, a smaller value of σ₀ was obtained in this work than the values reported by the previous authors.     

Comparison of Calculated Values with Measured Values on the Amount of TRU and FP Nuclides Accumulated in Gadolinium Bearing PWR Spent Fuels

Destructive analyses for five spent fuel samples taken from a Gd bearing fuel assembly were done. The measured amounts of actinides of ^234-238U, ²³⁷Np, ^238-242p_u 241,242m,243Am ^242,244Cm, and fission products of ¹³⁴Cs and ¹⁵⁴Eu were used for evaluating the accuracy of calculation made by CASMO-MICBURN and ORIGEN-2 codes. The effect of Gd on the neutron spectrum was taken into account in the CASMO-MICBURN calculation.

The amounts of ²³⁵U, ²³⁹Pu and ²⁴¹Pu calculated by CASMO-MICBURN agreed well with the observed values within about 3%. On the other hand, the amounts obtained from ORIGEN-2 calculation showed lower values than those observed, especially by —12% in average in ²³⁵U for Gd₂0₃U0₂ fuel. The main cause of this large difference may be attributed to the effect of Gd on the neutron spectrum. The amounts of the other actinides by both calculation codes revealed no significant difference in nearly 10% except for ^242mAm, in which a large fluctuation among the samples was observed. About 10% difference between the measured values and the calculated values was also observed for ¹³⁴Cs, but the calculated values for ¹⁵⁴Eu showed a significant difference from measured values.     

Analyses of used BWR fuel assay data with the integrated burnup code system SWAT4.0

Criticality safety of the fuel debris from the Fukushima Daiichi Nuclear Power Plant is one of the most important issues, and the adoption of burnup credit is desired for criticality safety evaluation. To adopt the burnup credit, validation of the burnup calculation codes is required. Assay data of the used nuclear fuel irradiated by the Fukushima Daini Nuclear Power Plant Unit 2 are evaluated to validate the SWAT4.0 code for solving the BWR fuel burnup problem. The calculation results revealed that the number densities of many heavy nuclides and fission products show good agreement with the experimental data, except for those of ²³⁷Np, ²³⁸Pu, and samarium isotopes. These differences were considered to originate from inappropriate assumption of void fraction. Our results implied overestimation of the (n, γ) cross-section of ²³⁷Np in JENDL-4.0. The Calculation/Experiment – 1 (C/E–1) value did not depend on the type of fuel rod (UO₂ or UO₂–Gd₂O₃), which was similar to the case of PWR fuel. The differences in the number densities of ²³⁵U, ²³⁹Pu, ²⁴⁰Pu, ²⁴¹Pu, ¹⁴⁹Sm, and ¹⁵¹Sm have a large impact on k_eff. However, the reactivity uncertainty related to the burnup analysis was less than 3%. These results indicate that SWAT4.0 appropriately analyzes the isotopic composition of BWR fuel, and it has sufficient accuracy to be adopted in the burnup credit evaluation of fuel debris.     

Some aspects of risk reduction strategy by multiple recycling in fast burner reactors of the plutonium and minor actinide inventories

This paper shows the impact of recycling light water reactor (LWR) mixed oxide (MOX) fuel in a fast burner reactor on the plutonium (Pu) and minor actinide (MA) inventories and on the related radioactivities. Reprocessing of the targets for multiple recycling will become increasingly difficult as the burnup increases. Multiple recycling of Pu + MA in fast reactors is a feasible option which has to be studied very carefully: the Pu (except the isotopes Pu-238 and Pu-240), Am and Np levels decrease as a function of the recycle number, while the Cm-244 level accumulates and gradually transforms into Cm-245. Long cooling times (10 + 2 years) are necessary with aqueous processing. The paper discusses the problems associated with multiple reprocessing of highly active fuel types and particularly the impact of Pu-238, Am-241 and Cm-244 on the fuel cycle operations. The calculations were performed with the zero-dimensional ORIGEN-2 code. The validity of the results depends on that of the code and its cross-section library. The time span to reduce the initial inventory of Pu + MA by a factor of 10 amounts to 255 years when average burnups are limited to 150 GW · d t⁻¹ (tonne).     

Development of laboratory standards for AMS measurement of 237Np

A new method for the preparation of ²³⁷Np/²³⁸U laboratory standards for accelerator mass spectrometry (AMS) measurement was developed at China Institute of Atomic Energy (CIAE). ²³⁷Np was generated by β-decay (6.75 days) of ²³⁷U produced via fast neutron induced ²³⁸U(n,2n)²³⁷U reaction. The ⁵⁸Co(n,2n)⁵⁹Co and ⁹³Nb(n,2n)^92mNb reactions were used to monitor the integral incident neutron flux. The cross-section of ²³⁸U(n,2n)²³⁷U was determined by measuring the emission rate of 208.0 keV γ-rays from ²³⁷U decay with a calibrated HPGe γ-ray spectrometer at CIAE. In order to correct for the self-attenuation of 208.0 keV γ-ray in the U sample, a ¹⁷⁷Lu–U method was introduced. The new correction method provided effective and high quality measurements of the cross-section for the ²³⁸U(n,2n)²³⁷U reaction; and the ²³⁷Np/²³⁸U ratio. A series of AMS standards for ²³⁷Np/²³⁸U (10^?9–10^?13) can be obtained by successive dilution of the original standard sample.     

Absolute Fission-Rate Distributions in Lithium and Hybrid Fusion Blanket Assemblies, (II)

The absolute fission rates of ²³⁵U, ²³⁷Np, ²³⁸U and ²³²Th were measured in four types of spherical blanket assemblies containing lithium and/or natural uranium and/or graphite. The results of measurement are compared with those of one-dimensional transport calculations employing 100-group neutron cross-sections obtained from the ENDF/B-IV data file. It is shown that the ratios between calculated and experimental values of ²³²Th, ²³⁸U and ²³⁷Np fission rates decrease with distance from the assembly center, where D-T neutrons are generated. An overestimation of about 50% is observed in the calculated ²³⁵U fission rate for the graphite reflector region.

One of the main sources of the disagreement proves to lie in the inability of the codes adopted for generating the multi-group cross-section to take account of the angular distributions of the secondary neutrons emanating from nonelastic reactions. The results of the analysis indicate that the method of calculation currently employed in fusion reactor neutronics overestimates the reflection of neutrons and underestimates the penetration of fast neutrons when a graphite reflector is used.     

Measurements of Thermal Neutron Cross Section and Resonance Integral for 237Np(n, γ)238Np Reaction

Making use of a standard neutron spectrum field with a pure Maxwellian distribution, the thermal neutron cross section for the ²³⁷Np(n, γ)²³⁸Np reaction was measured at a neutron energy of 0.0253 eV by the activation method. The result is 158±3 b, which is obtained relative to the reference value of 98.65±0.09 b for the ¹⁹⁷Au(n, γ)¹⁹⁸Au reaction. Although the data in JENDL-3 is larger by about 15% than the present value, the recently revised data in JENDL-3.2 is close to the present. The ENDF/B-V, ENDF/B-VI, JEF-2 and Mughabghab's data are also larger by 7–15%. Old measurements are larger by 7–18% than the present data.

The resonance integral for the ²³⁷Np(n, γ)²³⁸Np reaction was also measured relative to the reference value of 1,550±28 b for the ¹⁹⁷Au(n, γ)¹⁹⁸Au reaction with a 1/E standard neutron spectrum field. By defining the Cd cut-off energy as 0.5 eV for the ²³⁷Np(n, γ)²³⁸Np reaction, the present resonance integral is 652 ± 24 b, which is in good agreement with the JENDL-3, -3.2, ENDF/B-V, -VI, JEF-2 and Mughabghab's data. However, most of the old experimental data are, in general, larger by 24–38% than the present measurement.