Dependence of the Parameters of an Electronuclear System on the Composition of Mixed Uranium–Plutonium Fuel

Mathematical simulation is used to study the dependence of the parameters of an electronuclear system with uranium–plutonium mixed fuel on the amount of plutonium and ²⁴⁰Pu admixture in the fuel. As an example, a 10–20 kW(t) experimental system built at Dubna on the basis of a synchrocyclotron with 660 MeV protons is examined. A 2% admixture of ²⁴⁰Pu decreases the neutron multiplication factor from 0.95 to 0.9 and decreases the neutron yield and heat released by almost a factor of 2. This decrease can be compensated by increasing the plutonium content in the mixed fuel from 25 to 27%.     

Isotopic composition of plutonium in the bottom deposits of the Norwegian Sea and Greenland Sea and identification of the sources of contaminationhearder

The results of measurements of the isotopic atomic ratios²³⁸Pu/²³⁹Pu,²⁴⁰Pu/²³⁹Pu,²⁴¹Pu/²³⁹Pu,²⁴²Pu/²³⁹Pu in samples of bottom deposits are presented. The role of the main sources of plutonium contamination of the region is examined. It is shown that the contribution of the α activity of industrial plutonium (^238–240Pu) in bottom sediments from the European reprocessing plants, mainly, Sellafield (Great Britain), on the background due to global fallouts is 20 to 55% of the total α activity and 70–95% for the β activity (²⁴¹Pu). No escape of reactor and weapons plutonium from the nuclear powered submarine “Komsomolets” into the environment was observed. The lower limit of sensitivity of the method is 10⁻¹⁴g of plutonium in a sample, which corresponds to a plutonium α activity of 0.02 mBq in a 10 g sample of bottom deposits, 4 tables, 26 references. Translated from Atomnaya énergiya, Vol. 87, No. 4, pp. 286–294. October, 1999.     

Determination of the plutonium mass in containers with plutonium dioxide according to the γ radiation from spontaneous-fission products

A nondestructive method of determining the mass of large plutonium samples which is based on measuring their characteristic γ radiation is presented. The mass and the isotopic composition of the plutonium were determined according to a single measured γ spectrum in two energy ranges: the middle range, where the isotopic composition of plutonium is determined from the lines of the plutonium isotopes, and high-energy, where the radiation of the products of the spontaneous fission of plutonium gives information about the mass of the sample. The dependence of the counting rate in the peaks of the fission products on the effective mass of ²⁴⁰Pu was calibrated according to measurements performed with standard samples of the enterprise. As a result of the measurements, corrections were made for the self-absorption of γ radiation and induced fission, which were calculated by the Monte Carlo method. The error in determining the plutonium dioxide mass is 3–10% for containers with different cool-down times and isotopic composition of plutonium.     

Nondestructive determination of plutonium by gamma spectrometry and neutron well coincidence counting

Different NDA methods based on gamma ray and neutron measurements developed for the determination of Pu in solid samples is reported. In the gamma spectrometric measurements for solid samples, a method which takes advantage of the multiple γ-rays emitted by Pu and relies on the empirical relation between apparent mass of the sample and γ-ray energy was used. The method is applicable for the determination of small quantities of plutonium samples of non standard geometry by gamma ray spectrometry. The gross and coincidence neutron count rates for two different sets of standard Pu oxide powder samples were found to fall on different calibration lines. Isotopic composition of the two sets were determined using gamma ray spectrometry to obtain effective ²⁴⁰Pu content in the samples. A common calibration curve could be obtained when coincidence count rates were plotted vs. effective ²⁴⁰Pu content.     

Neutron-coincidence determination of plutonium taking into account multiplication and the actual multiplicity of leakage neutrons

Conclusions The model developed for describing the response of neutron-coincidence setups taking into account the real leakage-neutron multiplicity distribution (RMD approximation) is effective for analyzing samples of metallic plutonium with different isotopic composition, plutonium dioxide, and composites of mixed uranium — plutonium fuel in the form of powder and granules. The only inconvenience of this model, just as the BEH approximation, is that computational or experimental estimates of the multiplication effect must be obtained. The quantitiesK _m,S _s ^* , andS _in ^* can be calculated on personal computers. For IBM-PS/2 computers one calculation requires 5 min. In the future we intend to extend the approach developed in this paper to setups employing analysis of multiple neutron coincidences (triple, and so on). Then there is evidently no need to estimate the multiplication effect in advance. A. A. Bochvar All-Union Scientific-Research Institute of Standardization in Machine Building. FéI. Scientific-Research Institute of Nuclear Reactors. Translated from Atomnaya énergiya, Vol. 76, No. 1, pp. 47–54, January, 1994.     

采用无源符合中子法测量含钚材料

研究了无源符合中子法测量不同含Ｐｕ材料的方法和技术。采用高计数率符合中子计数器测量材料中的２４０Ｐｕ等核素自发裂变产生的裂变中子，通过计算等效２４０Ｐｕ量来确定材料中的总含Ｐｕ量。在现场实测了含Ｐｕ炉渣等十多类各种物理形态的材料。讨论了测量数据的修正及误差来源。测量误差一般在８％—１１％，大部分结果与天平称重估算法在１０％内符合。     

α-Rossi determination of deeply subcritical states of multiplying media

A method is developed to determine deeply subcritical states of multiplying media. The method is based on a combination of the α-Rossi method and the multiplying source method with a calculation of the corresponding isotopic-spatial correlation factor. The effective neutron multiplication coefficients in plutonium blocks are measured. The range of determination of the effective multipication coefficients was 0.3–0.45. The measured effective neutron multiplication coefficients in plutonium blocks agree well with the computational results. The effective neutron multiplication coefficients in the MATR subcritical assembly with VVéR fuel elements were measured for different water levels. The range of determination of the effective multiplication coefficients was 0.7–0.97. The measured effective neutron multiplication coefficients agree with the results obtained using the shooting-source method. __________ Translated from Atomnaya énergiya, Vol. 101, No. 2, pp. 140–148, August, 2006.     

Critical Experiments on Light-Water Moderated PuO2-UO2 Lattices

A series of critical experiments using mixed-oxide (PuO₂-UO₂) plutonium fuels was carried out at the Tank type Critical Assembly (TCA) in the Japan Atomic Energy Research Institute in cooperation with the Power Reactor and Nuclear Fuel Development Corporation. Isotopie compositions of the Pu were 68, 22, 7 and 2w/0 for ²³⁹Pu, ²⁴⁰Pu, ²⁴¹Pu and ²⁴²Pu, respectively, and the U was natural uranium.

Critical mass, power distribution and neutron activation distribution of Au wire were measured for five kinds of lattices in which the ratios of atomic density of H to Pu were 295, 402, 494, 704 and 922, respectively. The codes LASER, UGMG42-THERMOS and GTB-2 were used for obtaining three neutron group constants and the PDQ-5 code was used in two-dimensional diffusion calculations for the lattices.

The maximum differences of the calculated multiplication factors from the measured values distributed from –1.32 to 1.72%δk/k. Calculated thermal neutron density, epithermal neutron flux and power distributions showed good agreement with measured data.     

Application of Fusion Neutron Source for Denaturing of Plutonium

Potential of DT fusion neutron source to enhance proliferation resistance properties of plutonium by means of its isotopic denaturing is addressed. The approach is exemplified by denaturing of pure ²³⁹Pu and plutonium of typical LWR spent fuel through transmutation of neptunium. The essential feature of a fusion driven system proposed in the study is a zero mass balance of plutonium: total plutonium inventory is constant during irradiation. The system is capable to convert pure ²³⁹Pu into plutonium composition with more than 20% fraction of key ²³⁸Pu isotope during 1,000 d of irradiation under initial neutron loading of 1 MW.m^?2. Denaturing of LWR spent fuel plutonium under the same conditions would increase its ²³⁸Pu content up to 10-12%.     

Influence of the uncertainty of the microscopic cross sections of <Superscript>238</Superscript>U and plutonium isotopes on the neutron multiplication coefficient in VVÉR

The influence of the uncertainties in the microscopic cross sections of ²³⁸U and ^239–241Pu on the neutron multiplication coefficient in VVéR-1000 is studied. Data on the uncertainty of the cross sections were obtained by analyzing the ENDF/B files of the JENDL 3.3 system using the ERRORJ program. The characteristics of the reactor are calculated using the TRIFON 2.1 and SHERHAN programs. An approach associated with the calculation of the coefficients of sensitivity of the fuel assemblies and the reactor to standard samples in a small-group approximation is used. __________ Translated from Atomnaya énergiya, Vol. 103, No. 4, pp. 247–254, October, 2007.     

Energy Dependence of Plutonium Fission-Product Yields

A method is developed for interpolating between and/or extrapolating from two pre-neutron-emission first-chance mass-asymmetric fission-product yield curves. Measured ²⁴⁰Pu spontaneous fission and thermal-neutron-induced fission of ²³⁹Pu fission-product yields (FPY) are extrapolated to give predictions for the energy dependence of the n + ²³⁹Pu FPY for incident neutron energies from 0 to 16 MeV. After the inclusion of corrections associated with mass-symmetric fission, prompt-neutron emission, and multi-chance fission, model calculated FPY are compared to data and the ENDF/B-VII.1 evaluation. The ability of the model to reproduce the energy dependence of the ENDF/B-VII.1 evaluation suggests that plutonium fission mass distributions are not locked in near the fission barrier region, but are instead determined by the temperature and nuclear potential-energy surface at larger deformation.     

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.     

Effect of the neutron spectrum on the characteristics of Np, Am, and Cm transmutation chains

It is shown by analysis of the influence of the neutron spectrum on the neutron balance of the transmutation chains of ²³⁷Np and ²⁴¹Am that in a fast-reactor spectrum these chains are short and more neutrons are created (excess 0.8–1 neutron per starting nucleus) than absorbed in them. In a thermal-reactor spectrum, these chains are longer and more neutrons are absorbed (deficit 0.7–0.83 per starting nucleus) than created in them. Additional nuclear fuel, for example, ²³⁵U and ²³⁹Pu, must be burned to cover the neutron deficit. __________ Translated from Atomnaya énergiya, Vol. 100, No. 6, pp. 452–458, June, 2006.     

Application of active and passive neutron non destructive assay methods to concrete radioactive waste drums

This paper deals with the application of non-destructive neutron measurement methods to control and characterize 200 l radioactive waste drums filled with a concrete matrix. Due to its composition, and particularly to hydrogen, concrete penalizes the use of such methods to quantify uranium (U) and plutonium (Pu) components, which are mainly responsible of the α-activity of the waste. The determination of the alpha activity is the main objective of neutron measurements, in view to verify acceptance criteria in surface storage. Calibration experiments of the Active Neutron Interrogation (ANI) method lead to Detection Limit Masses (DLM) of about 1 mg of ²³⁹Pu_eff in the total counting mode, and of about 10 mg of ²³⁹Pu_eff in the coincidence counting mode, in case of a homogeneous Pu source and measurement times between one and two hours. Monte Carlo calculation results show a very satisfactory agreement between experimental values and calculated ones. Results of the application of passive and active neutron methods to control two real drums are presented in the last part of the paper. They show a good agreement between measured data and values declared by the waste producers. The main difficulties that had to be overcome are the low neutron signal in passive and active coincidence counting modes due to concrete, the analysis of the passive neutron signal in presence of ²⁴⁴Cm in the drum, which is a strong spontaneous fission neutron emitter, the variation of the active background with the concrete composition, and the analysis of the active prompt neutron signal due to the simultaneous presence of U and Pu in the drums.     

Improved Los Alamos model applied to the neutron induced fission of 239Pu and 240Pu and to the spontaneous fission of Pu isotopes

The Los Alamos model with multiple fission chances upgraded with (a) the linear relation between the average prompt gamma ray energy and the average prompt neutron multiplicity and (b) the dependence of the average fission fragment kinetic energy on the incident neutron energy, is used for the n+²³⁹Pu and n+²⁴⁰Pu reactions, and also for the spontaneous fission of ^237–241Pu isotopes. In the case of ²⁴⁰Pu fissioning nucleus the variation of the average energy released versus the incident neutron energy is also taken into account. The calculated prompt fission neutron spectra and average prompt neutron multiplicity well represent the experimental data, proving a better predictive power of the improved Los Alamos model.     

Experimental Study on the Importance of Accurate Neutron Angular Distribution Measurement for Determination of Neutron Yield in a 2.5 kJ Mather type PF

Three activation counters were constructed by thin wall Geiger and were placed in different angels with respect to anode axis; 0°, 45°, 90°; where the Geigers have 20 cm distance from anode top of the 2.5 kJ SBUPF1 plasma focus device. The counters were calibrated by a 5 Ci Am–Be neutron source with source removal method. A computer program receives the Results of each plasma focus experiment and background count rates, estimates the angular distribution of neutron emission and calculates the neutron yield of each shot. The neutron yield of the device was measured about 6 × 10⁷ neutrons per shot. The results, indicate that using three counters (instead of one counter in 0°) in different angles for determination of total neutron yield, gives a more accurate measurement (up to 28% is measured in sample shots), and this error is bigger in those shots that thermonuclear fusion mechanism have greater share in neutron production.     

Representation of the fission neutron spectrum of235U,239Pu, and252Cf and the reactor spectrum as a superposition of five inelastic scattering functions

The results of investigations of the representation of the fission neutron spectra of²³⁵U,²³⁹Pu, and²⁵²Cf and neutron spectra at the core center of fast reactors as a superposition of five Weisskopf functions are presented. This representation best reflects the physics of nuclear fission and shows that the spectra at the core center of fast reactors are reconstructed with an error close to the error of the best studied spectra in the worldwide practice of neutron measurements. 3 tables, 13 references. All-Russia Scientific-Research Institute of Physical-Technical and Radio Engineering Measurements. Translated from Atomnaya énergiya, Vol. 88, No. 4, pp. 292–299, April, 2000.     

Conditions of Transmutation of Plutonium, Americium, and Curium in Nuclear Facilities

The transmutation of plutonium, americium, and curium isotopes in reactors with thermal and fast neutron spectra is analyzed. The radiotoxicity of a nuclide mixture is calculated. For transmutation of plutonium the radiotoxicity is maximum at the initial stage of irradiation as a result of the production of ²⁴¹Pu and, after further irradiation, the production of ²⁴⁴Cm. For transmutation of americium and curium the maximum radiotoxicity is 2–2.5 times greater than the initial value; it is due to the formation of ²³⁸Pu. It is established that the neutron flux density affects the rate of decrease of radiotoxicity. 5 tables, 5 references.     

Radiochemical Determination of Plutonium Isotopes and Americium-241 in a Plutonium Sample

Isotopic ratios of ²³⁸Pu, ²³⁹Pu, ²⁴⁰Pu and ²⁴¹Am in a plutonium sample were approximately determined by means of combined α- and γ-ray spectrometry without chemical separation of americium from plutonium. The intensity of α-ray followed by internal conversion was determined by measuring the intensity of LX-rays. The ratio of ²⁴⁰Pu to ²³⁹Pu was obtained from the ratio between the α-ray intensity thus determined and the total α-ray intensity from the two nuclides.     

Burn up extension with slightly enriched uranium and plutonium fuel in CANDU reactors

A study of fuel burn-up and concentrations of uranium and plutonium isotopes for the three fuel cycles of a CANDU reactor are carried out in the present work. The infinite and effective multiplication factors are calculated as a function of fuel burn up for the natural UO₂ fuel, 1.2% enriched UO₂ fuel and for the 0.45% PuO₂-UO₂ fuel. The amount of ²³⁵U and ²³⁸U consumed and ²³⁹Pu, ²⁴⁰Pu and ²⁴¹Pu produced in the three fuel cycles are also calculated and compared.