A setup for active neutron analysis of the fissile material content in fuel assemblies of nuclear reactors

An active neutron method for measuring the residual mass of ²³⁵U in spent fuel assemblies (FAs) of the IRT MEPhI research reactor is presented. The special measuring stand design and uniform irradiation of the fuel with neutrons along the entire length of the active part of the FA provide high accuracy of determination of the residual ²³⁵U content. AmLi neutron sources yield a higher effect/background ratio than other types of sources and do not induce the fission of ²³⁸U. The proposed method of transfer of the isotope source in accordance with a given algorithm may be used in experiments where the studied object needs to be irradiated with a uniform fluence.     

Radiation effect of neutrons produced by D–D side reactions on a D–3He fusion reactor

One of the most important characteristics in D–³He fusion reactors is neutron production via D–D side reactions. The neutrons can activate structural material, degrading them and ultimately converting them into high-level radioactive waste, while it is really costly and difficult to remove them. The neutrons from a fusion reactor could also be used to make weapons-grade nuclear material, rendering such types of fusion reactors a serious proliferation hazard. A related problem is the presence of radioactive elements such as tritium in D–³He plasma, either as fuel for or as products of the nuclear reactions; substantial quantities of radioactive elements would not only pose a general health risk, but tritium in particular would also be another proliferation hazard. The problems of neutron radiation and radioactive element production are especially interconnected because both would result from the D–D side reaction. Therefore, the presentation approach for reducing neutrons via D–D nuclear side reactions in a D–³He fusion reactor is very important. For doing this research, energy losses and neutron power fraction in D–³He fusion reactors are investigated. Calculations show neutrons produced by the D–D nuclear side reaction could be reduced by changing to a more ³He-rich fuel mixture, but then the bremsstrahlung power loss fraction would increase in the D–³He fusion reactor.     

Physical particularities of nuclear reactors using heavy moderators of neutrons

In nuclear reactors, thermal neutron spectra are formed using moderators with small atomic weights. For fast reactors, inserting such moderators in the core may create problems since they efficiently decelerate the neutrons. In order to form an intermediate neutron spectrum, it is preferable to employ neutron moderators with sufficiently large atomic weights, using ²³³U as a fissile nuclide and ²³²Th and ²³¹Pa as fertile ones. The aim of the work is to investigate the properties of heavy neutron moderators and to assess their advantages. The analysis employs the JENDL-4.0 nuclear data library and the SCALE program package for simulating the variation of fuel composition caused by irradiation in the reactor. The following main results are obtained. By using heavy moderators with small neutron moderation steps, one is able to (1) increase the rate of resonance capture, so that the amount of fertile material in the fuel may be reduced while maintaining the breeding factor of the core; (2) use the vacant space for improving the fuel-element properties by adding inert, strong, and thermally conductive materials and by implementing dispersive fuel elements in which the fissile material is self-replenished and neutron multiplication remains stable during the process of fuel burnup; and (3) employ mixtures of different fertile materials with resonance capture cross sections in order to increase the resonance-lattice density and the probability of resonance neutron capture leading to formation of fissile material. The general conclusion is that, by forming an intermediate neutron spectrum with heavy neutron moderators, one can use the fuel more efficiently and improve nuclear safety.     

Operation of CANDU power reactor in thorium self-sufficient fuel cycle

This paper presents the results of calculations for CANDU reactor operation in thorium fuel cycle. Calculations are performed to estimate the feasibility of operation of heavy-water thermal neutron power reactor in self-sufficient thorium cycle. Parameters of active core and scheme of fuel reloading were considered to be the same as for standard operation in uranium cycle. Two modes of operations are discussed in the paper: mode of preliminary accumulation of ²³³U and mode of operation in self-sufficient cycle. For the mode of accumulation of ²³³U it was assumed for calculations that plutonium can be used as additional fissile material to provide neutrons for ²³³U production. Plutonium was placed in fuel channels, while ²³²Th was located in target channels. Maximum content of ²³³U in target channels was estimated to be ∼13 kg/t of ThO₂. This was achieved by irradiation for six years. The start of the reactor operation in the self-sufficient mode requires ²³³U content to be not less than 12 kg/t. For the mode of operation in self-sufficient cycle, it was assumed that all channels were loaded with identical fuel assemblies containing ThO₂ and certain amount of ²³³U. It is shown that nonuniform distribution of ²³³U in fuel assembly is preferable.      

A conceptual high flux reactor design with scope for use in ADS applications

A 100 MWt reactor design has been conceived to support flux level of the order of 10¹⁵ n/cm²/s in selected flux trap zones. The physics design considers high enriched metallic alloy fuel in the form of annular plates placed in a D₂O moderator tank in a hexagonal lattice arrangement. By choosing a tight lattice pitch in the central region and double the lattice pitch in the outer region, it is possible to have both high fast flux and thermal flux trap zones. By design the flux level in the seed fuel has been kept lower than in the high flux trap zones so that the burning rate of the seed is reduced. Another important objective of the design is to maximize the time interval of refueling. As against a typical refueling interval of a few weeks in such high flux reactor cores, it is desired to maximize this period to as much as six months or even one year. This is possible to achieve by eliminating the conventional control absorbers and replacing them with a suitable amount of fertile material loading in the reactor. Requisite number of seedless thorium-aluminum alloy plates are placed at regular lattice locations vacated by seed fuel in alternate fuel layers. It is seen that these thorium plates are capable of acquiring asymptotic fissile content of 14 g/kg in about 100 days of irradiation at a flux level of 8 × 10¹⁴ n/cm²/s. In summary, the core has a relatively higher fast flux in the central region and high thermal flux in the outer region. The present physics design envisages a flat core excess reactivity for the longest possible cycle length of 6 months to one year. It is also possible to modify the design for constant subcriticality for about the same period or longer duration by considering neutron spallation source at the centre and curtailing the power density in the inner core region by shielding it with a layer of thoria fuel loading.      

Activation Characteristics of Fuel Breeding Blanket Module in Fusion Driven Subcritical System

Shortage of energy resources and production of long-lived radioactivity wastes from fission reactors are among the main problems which will be faced in the world in the near future. The conceptual design of a fusion driven subcritical system (FDS) is underway in Institute of Plasma Physics, Chinese Academy of Sciences. There are alternative designs for multi-functional blanket modules of the FDS, such as fuel breeding blanket module (FBB) to produce fuels for fission reactors, tritium breeding blanket module to produce the fuel, i.e. tritium, for fusion reactor and waste transmutation blanket module to try to permanently dispose of long-lived radioactivity wastes from fission reactors, etc. Activation of the fuel breeding blanket of the fusion driven subcritical system (FDSFBB) by D-T fusion neutrons from the plasma and fission neutrons from the hybrid blanket are calculated and analysed under the neutron wall loading 0.5 MW/m2 and neutron fluence 15 MW.yr/m2. The neutron spectrum is calculated with the worldwide-used transport code MCNP/4C and activation calculations are carried out with the well known European inventory code FISPACT/99 with the latest released IAEA Fusion Evaluated Nuclear Data Library FENDL-2.0 and the ENDF/B-V uranium evaluated data. Induced radioactivities, dose rates and afterheats, etc, for different components of the FDS-FBB are compared and analysed.     

Program of Fundamental-Interaction Research for the Ultracold-Neutron Source at the the WWR-M Reactor

The use of ultracold neutrons opens unique possibilities for studying fundamental interactions in particles physics. Searches for the neutron electric dipole moment are aimed at testing models of CP violation. A precise measurement of the neutron lifetime is of paramount importance for cosmology and astrophysics. Considerable advances in these realms can be made with the aid of a new ultracold-neutron (UCN) supersource presently under construction at Petersburg Nuclear Physics Institute. With this source, it would be possible to obtain an UCN density approximately 100 times as high as that at currently the best UCN source at the high-flux reactor of the Institute Laue–Langevin (ILL, Grenoble, France). To date, the design and basic elements of the source have been prepared, tests of a full-scale source model have been performed, and the research program has been developed. It is planned to improve accuracy in measuring the neutron electric dipole moment by one order of magnitude to a level of 10^?27 to 10^?28e cm. This is of crucial importance for particle physics. The accuracy in measuring the neutron lifetime can also be improved by one order of magnitude. Finally, experiments that would seek neutron–antineutron oscillations by employing ultracold neutrons will become possible upon reaching an UCN density of 10³ to 10⁴ cm^?3. The current status of the source and the proposed research program are discussed.     

Fuel cycle for a fusion neutron source

The concept of a tokamak-based stationary fusion neutron source (FNS) for scientific research (neutron diffraction, etc.), tests of structural materials for future fusion reactors, nuclear waste transmutation, fission reactor fuel production, and control of subcritical nuclear systems (fusion–fission hybrid reactor) is being developed in Russia. The fuel cycle system is one of the most important systems of FNS that provides circulation and reprocessing of the deuterium–tritium fuel mixture in all fusion reactor systems: the vacuum chamber, neutral injection system, cryogenic pumps, tritium purification system, separation system, storage system, and tritium-breeding blanket. The existing technologies need to be significantly upgraded since the engineering solutions adopted in the ITER project can be only partially used in the FNS (considering the capacity factor higher than 0.3, tritium flow up to 200 m³Pa/s, and temperature of reactor elements up to 650°C). The deuterium–tritium fuel cycle of the stationary FNS is considered. The TC-FNS computer code developed for estimating the tritium distribution in the systems of FNS is described. The code calculates tritium flows and inventory in tokamak systems (vacuum chamber, cryogenic pumps, neutral injection system, fuel mixture purification system, isotope separation system, tritium storage system) and takes into account tritium loss in the fuel cycle due to thermonuclear burnup and β decay. For the two facility versions considered, FNS-ST and DEMO-FNS, the amount of fuel mixture needed for uninterrupted operation of all fuel cycle systems is 0.9 and 1.4 kg, consequently, and the tritium consumption is 0.3 and 1.8 kg per year, including 35 and 55 g/yr, respectively, due to tritium decay.     

Angular distributions in quasi-fission reactions: Evidence for incomplete relaxation of the tilting mode

The integral β^?-spectra of²³⁵U and²³⁹Pu fission products have been measured with a plastic scintillator telescope at an external neutron guide tube at the high flux reactor of the ILL in Grenoble. The highly enriched targets (150 – 800 γg/cm²) were placed in a fission chamber at a distance of approximately 110 m from the reactor core. From the measured beta-spectra absolute counting rates per MeV and fission have been calculated, which are compared with the results of earlier experiments of other authors and with recent theoretical calculations.     

Estimation of the ultracold neutron production by a source designed for the WWR-M reactor

The results of calculations related to designing an ultracold neutron source with superfluid helium for the WWR-M reactor have been presented. The ultracold neutron production rate has been estimated for different types of premoderator chamber filling. The dependence of this rate on the temperature of helium has been determined. If the premoderator chamber is filled with liquid orthodeuterium, the ultracold neutron production rate remains almost constant in the range of helium temperatures of 1.0–1.5 K and is as high as 3.1 × 10³ cm^–3 s^–1.     

The intense neutron generator of 14 MeV neutrons

An innovative intense neutron generator of 14 MeV neutrons for the irradiation of future reactor materials is presented. Negative pions are produced inside a 5–10 T magnetic field by an intense deuteron beam interacting with a carbon target. The pions and the muons from pion decay in flight are collected in the backward direction and stopped in a deuterium-tritium-hydrogen target of high density. Using an 18 MW deuteron beam at 1.5 GeV (12 mA=7.5 × 10¹⁶d/s), circa 10¹⁶gt^– /s can be generated, decaying to muons of which up to 10¹⁵ µ^–/s stop in the D/T/H mixture. Assuming X_c=100 fusions per muon, the µCF source produces 14 MeV neutrons with a source strength of up to 10¹⁷ n/s, i.e. a neutron power of 200 kW. The environment of the second target, the neutron source itself, can be made to resemble part of the Tokamak ring to be simulated for irradiation test samples.     

Design of a graphite-moderated 241Am–Li neutron field to simulate reactor spectra

A neutron calibration field using ²⁴¹Am–Li sources and a moderator was designed to simulate the neutron fields found outside a reactor. The moderating assembly selected for the design calculation consists of a cube of graphite blocks with dimensions of 50 cm by 50 cm by 50 cm, in which the ²⁴¹Am–Li sources are placed. Monte Carlo calculations revealed the optimal depth of the source to be 15 cm. This moderated neutron source can be used to provide a test field that has a large number of intermediate energy neutrons with a small portion of MeV component.     

ADS瞬态分析中的概率论-确定论耦合方法

针对加速器驱动次临界系统(ADS)瞬态问题,采用预估校正改进准静态方法(PCQS)处理时空中子动力学方程中的时间自变量,采用蒙特卡罗方法处理相应的空间-角度-能量自变量,重点解决了低次临界度下模拟计算不稳定的问题,验证了TWGIL-Seed-Blanket动力学基准问题和小型模拟ADS问题,得到瞬态过程的功率变化结果,与基于其他方法的程序比较,经初步验证取得了较好结果,证明了该耦合方法可行。     

Measurements of Fission and Radioactive Capture Reaction Rates Inside the Fuel of the Ipen/MB-01

This work presents the measures of the nuclear reaction rates along the radial direction of the fuel pellet by irradiation and posterior gamma spectrometry of a thin slice of fuel pellet of UO₂ at 4.3% enrichment. From its irradiation, the rate of radioactive capture and fission had been measured as a function of the radius of the pellet disk using a Ortec GMX HPGe detector. Lead collimators had been used for this purpose. Simulating the fuel pellet in the pin fuel of the IPEN/MB-01 reactor, a thin UO₂ disk is used, being inserted in the interior of a dismountable fuel rod. This fuel rod is then placed in the central position of the IPEN/MB-01 reactor core and irradiated during 1 h under a neutron flux of 5 ×10⁸ n/cm² s. In gamma spectrometry, 10 collimators with different diameters have been used; consequently, the nuclear reactions of radioactive capture that occurs in atoms of ²³⁸U and the fission that occurs on both ²³⁵U and ²³⁸U are measured in function of 10 different regions (diameter of collimator) of the UO₂ fuel pellet disk. Nuclear fission produces different fission products such as ¹⁴³Ce with a yield fission of 5.9% which decay is monitored in this work. Corrections in geometric efficiency due to introduction of collimators on HPGe detection system were estimated using photon transport of MCNP-4C code. Some calculated values of nuclear reaction rate of radioactive capture and fission along the radial direction of the fuel pellet obtained by Monte Carlo methodology, using the MCNP-4C code, are presented and compared to the experimental data showing very good agreement.     

New measurements of the neutron electric dipole moment

We report a new measurement of the neutron electric dipole moment with the PNPI EDM spectrometer using the ultracold neutron source PF2 at the research reactor of the ILL. Its first results can be interpreted as a limit on the neutron electric dipole moment of |d _n | < 5.5 × 10^?26 e cm (90% confidence level).     

Gamma rays from thermal neutron capture in gold

The thermal column of the 10 kW University of Maryland pool reactor was used as the neutron source for this nuclear spectroscopy study. A 40 cm³ lithium drifted germanium crystal was the gamma (γ) ray detector used to measure the high energy prompt photon emissions from thermal neutron capture. The energy region was from 2.6–7.0 MeV. New capture gamma ray lines were observed.     

Analysis of some Elements in Fuel Oil by using Neutron Capture Gamma Ray Spectroscopy

The analysis of some elements in fuel oil was done by using neutron capture gamma ray spectroscopy. The gamma rays emitted after the reaction of (n, γ) from the fuel oil atoms bombarded with the thermal neutrons emitted from a ²⁵²Cf source were investigated in the energy interval 0 … 10 MeV. The impurity elements in the sample were determined.     

Absence of voids in fast neutron irradiated zirconium by small angle neutron scattering

Abstract

The aim of this research was to resolve a difference of opinion in the literature on the presence of voids in fast neutron irradiated zirconium. There is a great interest in the study of zirconium, since zirconium and its alloys are used extensively in modern power reactors, for example in the fuel rods as a containment material for enriched uranium. A polycrystalline sample of zirconium was irradiated in the HERALD reactor at 40°C with 10²⁰ fast neutrons per cm^?2. The neutron scattering from irradiated and unirradiated standard samples was studied over a wide Q range from 0.001 to 1.12 Å^?1 on a D11 Spectrometer at the ILL (France). The defect cross-section (the difference between the scattering of the standard zirconium crystal and irradiated crystal) was nearly flat as a function of Q (momentum transfer vector) with an average value of 8.5 mb/Str/atom. This indicates a point defect concentration of about 1.8%. Thus the absence of any small angle (Q dependent) defect scattering indicates that large damage regions (e.g. voids) are not produced in zirconium by fast neutron irradiation.     

Nuclear pumping of a3He-Cs metal-vapor mixture

Results of a spectroscopic study of low temperature atmospheric density³He-Cs plasmas created by the³He(n, p)³T reaction at a thermal neutron flux of about 8×10¹¹ n/cm² s are presented. Measurements were carried out using the steady state nuclear reactor of the Moscow Engineering Physics Institute as a neutron source in the temperature range 70–450°C. The possibility of direct nuclear pumped³He-Cs laser action is indicated by calculations ofnp,ns CsI level populations.     

On the use of averaging stopping power of recoiled nuclei for determining maximum neutron energy of 241Am–Be by superheated drop detectors

The average stopping power of the recoiled nuclei generated by neutron elastic interactions with the Freon-12 drops in a superheated drop detector has been used to determine the maximum neutron energy of the ²⁴¹Am–Be source. In an elastic interaction of neutrons with the Freon-12 liquid, the nuclei of ¹²C, ¹⁹F and ³⁵Cl with different values of stopping power are scattered. The stopping power of these scattered nuclei corresponding to the energy transferred to them through the head-on collision was extracted from the SRIM code. The stopping power values were weighted by considering the neutron–nucleus elastic scattering cross section and the number of each nucleus in the Freon-12 molecule and the average stopping power was calculated from known neutron energy.The maximum energy of the ²⁴¹Am–Be neutron source was estimated as 10.9 ± 3.0 MeV. The consistency between the determined energy and the other reported values confirms the validity of using the average stopping power in the superheated drop detectors. The average stopping power was also used to determine the threshold neutron energy as a function of external applied pressure at different temperatures. Knowing the threshold neutron energy as function of applied pressure, can be used in pressure scanning method for neutron spectrometry by superheated drop detectors.