<Emphasis Type="Italic">N</Emphasis>-spline reconstruction of neutron spectra on the basis of activation measurements

A new approach to reconstructing the neutron spectra of reactor facilities on the basis of activation measurements is proposed and investigated. A new parameterization of a neutron spectrum is introduced; this representation is in the form of a special “neutron” spline. A generalized algorithm for minimizing the directed deviation is used for the reconstruction. To analyze the possibilities of the approach developed, the spectra of the BARS-5, IGRIK, and YaGUAR reactor facilities are reconstructed on the basis of activation measurements. The reconstructed spectra agree well with previous results. __________ Translated from Atomnaya énergiya, Vol. 104, No. 5, pp. 295–302, May, 2008.     

Research-reactor neutron fields for calibrating in-reactor detectors in nuclear power plants

The neutron spectra of research and power reactors are compared. The spectra were measured by the neutron-activation method and calculated using the KASKAD computer code. The a priori spectrum in the calculation was constructed as a superposition of physically validated spectra. A method of calibrating in-reactor detectors in nuclear power plants on the basis of the sensitivity to the ²³⁵U fission rate in 1 g of uranium using the neutron fields of research reactors is proposed. __________ Translated from Atomnaya énergiya, Vol. 100, No. 2, pp. 97–107, February, 2006.     

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.     

Measurement of neutron spectra in the thermal column of the reactor of a power station

Knowledge of neutron spectra In nuclear reactors allows comparison of various theories of the slowing down of neutrons with experiment, and also allows carrying out reactor calculations which are based on actual neutron distributions therein. In this paper is described a neutron intensity monochtomator Intended for the measurement of neutron spectra in the energy interval 0 to 0.5 ev.Results are given for measurements for neutron spectra in the thermal column of the reactor of an atomic power station. Discontinuities in the neutron flux were discovered at neutron velocities of 600, 1000 and 1650 m/sec; an analysis is given of the causes of discontinuities of the neutron flux; an evaluation is made of the inelastic scattering cross section for neutrons in graphite. By the method of least squares, the temperature of the neutron gas was found, being equal to 354 ° K at a graphite temperature of 304 ° K.In conclusion we consider it our duty to express gratitude to A. K. Krasin and B. G. Dubovskii for interest and help in the work and F. L. shapiro for valuable interpretation of previous results.     

The use of resonance detectors for the investigation of neutron spectra in fast-neutron reactors

The possibility of the investigation of the low energy portion of the neutron spectra in reflecting fast reactors by activated resonance detectors is considered.Absorber difference and 1/v absorption methods are illustrated by an example of the measurement of the flux distribution of resonance neutrons with energies of 4.9 ev (Au¹⁹⁷) and 2.95 kev (Na²³) in the reflecting reactors BR-1 and BR-5. It is shown that the neutron spectrum region from one to several thousand electron volts can be studied in adequate detail with the aid of the set of detectors described.The authors express deep thanks to A. I. Leipunskii for interest in the work and to I. I. Bondarenko and V. V. Orlov for helpful advice and comments.     

Application of serpan's damage function to heavy-water moderated power reactors and its control with surveillance data of the karlsruhe power reactor,MZFR

For calculating $\text{ΔNDT}$, the irradiation-induced shift of the nil ductility transition temperature, for the design of pressurized water reactor vessels, literature data envelope curves are used. In the case of light-water moderated reactors no correction of different spectra influence is considered. Calculations with the damage functions, published by Serpan et al., demonstrate that neglecting spectrum influence yields conservative design data. For heavy-water moderated reactors an adjusting factor is calculated from Serpan's damage function. The $\text{ΔNDT}$ measured with samples of the ASTM reference steel A302-B, irradiated in the heavy-water reactor, MZFR, is compared to irradiation results of samples from the same steel plate irradiated in the Oak Ridge low intensity test reactor (LITR) which are published by NRL. From this comparison an experimental adjusting factor is derived. Both calculated damage function and experimental factors agree very well. This result demonstrates that Serpan's damage function is applicable also to MZFR. surveillance samples. The neutron spectrum in the MZFR surveillance location is extremely soft. Thermal neutron population is 900 times the fast flux ($\text{E > 0.9}\text{MeV}$). Its damage contribution, calculated with damage function, amounts to 73%.     

计算机模拟γ测量谱

一个建立在微机DOS环境下的计算机模拟γ射线测量谱系统以一个实测的^135Cs谱作为模板,对活化测量谱和同位素源的γ射线谱进行了成形模拟。谱中包含了对全能峰、康普顿沿、康普顿坪和反散射峰的模拟。模拟系统具有谱峰面积计算、谱成形、谱显示、谱处理、谱存盘等功能。利用该模拟程序对^137Cs源、^152Eu源、^60Co源的γ射线谱以及锆样品和岩石样品的中子活化测量谱进行了模拟,模拟谱与实际测量谱非常接近。     

Measurement of the subcriticality of power-stressed reactors with an analog reactimeter

Conclusions A technique has been developed for measuring the effects of reactivity in a subcritical reactor with an analog reactimeter. It is based on the compensation of the current applied to the reactimeter input from the neutron detector. The compensation of current produced by neutrons of the subcritical multiplying assembly formalizes the algorithm for reactivity calculation, making it an adequate model of a reactor with a source and making it possible to determine the subcriticality without prior entry into the critical state. In this case the measurements are made in the presence of neutron sources characteristic of power-stressed reactors. The regular devices of the control and safety system could be used to produce unsteady variation of the neutron flux.All of this permits the proposed method to be extended to zero-power reactors and to power-stressed reactors. Once the subcriticality has been measured an analog of the neutron source is introduced into the reactimeter. This instrument measures the effects of reactivity in the subcritical state without the reactor being previously put into the critical state, monitors the entry of the reactor into the critical state by checking the reactivity, and makes all the measurements usually made with analog reactimeters. If the intensity of the source does not change during measurements (5–10 min) when chambers sensitive to rays (e.g., KNK-56 chambers) are used as neutron-flux detectors, then the accuracy of -ray compensation does not affect the results of the measurements.Translated from Atomnaya Énergiya, Vol. 45, No. 5, pp. 375–376, November, 1978.     

Measurement of neutron spectra in uranium-water and uranium-monoisoprolyldiphenyl lattices

The neutron spectra in a cell of a heterogeneous subcritical system are investigated in detail. The neutron spectra were measured by means of a mechanical velocity selector and a time analyzer. A water-cooled water-moderated reactor (WWR) was used as the neutron source for the subcritical system. The results indicate that the neutron spectrum differs from the Maxwell equilibrium distribution both in the uranium and in the moderator. The neutron spectrum changes sharply at the boundary between the uranium block and the moderator. The spectrum varies only slightly within the block and within the moderator. A simple semiempirical functional relationship is found to exist between the thermal neutron spectra in the moderator and in the block.Translated from Atomnaya Énergiya, Vol. 13, No. 6, pp. 547–555, December, 1962     

New 6Li Sandwich Counter for Measurement of Neutron Spectra in Fast Neutron Systems

A new form of ⁶Li sandwich counter has been developed for use in measurements of neutron spectra in fast critical assemblies. The counters have two large surface area diodes approached very close to each other, on one of which is evaporated the target ⁶Li. This arrangement proves to provide a ten-fold increase of detection efficiency, while the resolution is improved by approximately ½ compared with currently used counters. The new counters were first tested on mono-energetic neutrons, after which neutron spectrum measurements were performed on natural uranium blocks, in the fast source reactor “YAYOI”, in the FCA IV-l-P and in the FCA V-3 core. The resulting spectra are compared with the measurements made with other types of counter and with calculated data. Proof was obtained that the new counter can serve practically and usefully for in-core measurements of fast neutron energy spectrum between 0.8 and 5 MeV.     

活化法测量散裂靶中子能谱的实验验证

散裂靶中子的能谱对加速器驱动次临界系统的倍增因数和嬗变率等影响很大,计算表明散裂靶中子谱在MeV能区与裂变中子谱相近。本文利用活化法测量临界装置的泄漏中子谱和中子注量率,提出了用In、Al、Mg、Ti、Au、Zn、Ni、Rh、Fe和Co等活化箔测量散裂靶中子能谱和中子注量率的方案。结果表明,将活化箔在散裂靶中子场中辐照5h,中子注量最高达5×1014 cm-2量级,辐照后1h内取出活化箔,根据半衰期的长短安排测量顺序,可测量散裂靶的中子能谱和中子注量率。     

Measurement and Analysis of Neutron Spectra in Lithium Fluoride and Polytetrafluoroethylene Piles

For the assessment of neutron cross section data for fluorine, angular neutron spectra in the lithium fluoride (LiF) and polytetrafluoroethylene ((CF₂)n) piles were measured in the energy range from a few keV to a few MeV by the time-of-flight method with an electron linac, and the results were compared with those calculated by using nuclear data from JENDL-2 and ENDF/B-IV. Spatial distributions of neutron and X-ray fluxes were also measured in the test piles by the activation method, and the influence of photoneutrons generated in the sample material on the neutron spectrum in each pile was estimated. As a result, it was found that their influence on the neutron spectrum shape below 1 MeV was not so large as was necessary to be taken into account for the present assessment.

The calculated spectra using the JENDL-2 data and the ENDF/B-IV data show generally good agreement with those measured in both piles. However, both calculations underestimate the neutron fluxes around several 100 keV, and overestimate those below 100 keV, when they are normalized in the energy range of 10 keV～1 MeV. Large discrepancies are found between the shapes of the measured and calculated spectra around the resonances of fluorine cross section below 100 keV. The present measurements and analyses suggest that the reevaluations of the inelastic and elastic scattering cross sections below 1MeV and the resonance cross sections below 100 keV are necessary to reduce the observed discrepancies.     

Neutronics diagnostics for European ITER TBMs: Activation foil spectrometer for short measurement cycles

An important aim of neutronics Test Blanket Module (TBM) experiments in ITER will be to check the prediction accuracy of nuclear responses in an environment closer to a future fusion power reactor than so far provided by existing facilities. The development of measurement methods suitable for the harsh environment in an ITER TBM has been addressed in several recent R&D programs supported by Euratom. Within this framework, KIT is developing an activation foil spectrometer for the measurement of local neutron flux densities in the TBM. We intend to establish a measurement method which allows to record the induced activities in small packages of activation foils simultaneously and to calculate the corresponding spectral neutron flux densities with moderate time resolution of tens of seconds immediately after extraction from the TBM. In the present work we propose a candidate set of activation foil materials which cover the neutron energy range from thermal to 14 MeV. In order to assess their basic suitability for such measurements, we have computed induced gamma-ray activities in the foils using a calculated neutron spectrum in a representative position in the European HCPB TBM assuming a short irradiation time of 30 s. In a further step we have investigated pulse height spectra which would be obtained in a typical gamma-ray measurement arrangement in a HPGe detector and concluded that the proposed set of activation foils should be basically suitable for such a measurement system but require improvement of relevant cross sections uncertainties.     

Prompt-Neutron Spectrum Formation Mechanism in 235U, 239Pu, and 252Cf Fission

The mechanism of the formation of the prompt-neutron spectrum in fissioning of ²³⁵U and ²³⁹Pu by thermal and fast neutrons and spontaneous fission of ²⁵²Cf is investigated. A method is proposed for the formation of the prompt-neutron spectrum in fissioning of nuclei as a superposition of three partial Weisskopf evaporation spectra with average neutron energy 0.4, 2.06, and 2.8 MeV.     

Azimuthal flux distribution measurements around fuel rods in reduced-moderation LWR lattices

An advanced analysis method named “micro reactor physics approach” was proposed, and the approach is needed for future reactor design considering the neutron behavior in fuel pellets. In order to validate the approach, neutron flux distribution measurements in a fuel pellet should be required. We have measured azimuthal flux distribution of fuel rods in Toshiba Nuclear Critical Assembly (NCA). A foil activation method with metallic foils was used for the measurement. Measured values were analyzed by a continuous energy Monte Carlo code MVP with the JENDL-3.3 library. The measurements are useful for the validation of an advanced fuel design method considering the neutron behavior in fuel pellets.     

Neutronics study for the installation of the LBVM in the medium flux area of IFMIF

IFMIF (International Fusion Materials Irradiation Facility) will be a fusion dedicated facility producing a large amount of neutrons with the appropriate energy spectrum to test materials and subcomponents for DEMO and future Fusion Power Plants.While the high flux area of IFMIF will be devoted to reduced activation structural materials for first wall and blanket, the medium flux area will be dedicated to functional materials for breeder blankets. In particular, the Liquid Breeder Validation Module (LBVM), will host experiments related with functional materials for liquid breeder blankets. Since IFMIF neutron spectra have been intended to fit the most irradiated areas of a fusion reactor in the high flux area, the irradiation conditions in the LBVM placed in the medium flux area of IFMIF have been assessed. The effect of some neutron shifter/reflector components to optimize the neutron spectra have been evaluated in order to find out the proper irradiation conditions for functional materials for liquid breeder blankets.Therefore, the objective of this report is to summarize the neutronic calculations developed to evaluate the viability of IFMIF neutron source to perform relevant irradiation experiments on functional materials for liquid breeder blanket concept for future nuclear fusion power reactors (ITER, DEMO). The irradiation parameters evaluated for this purpose are: the tritium production for liquid breeder material (Pb–17Li) and the damage dose (dpa) and gas production to damage dose ratios for Al₂O₃ and SiC functional materials.The main conclusion is that, it is possible to perform relevant irradiation experiments on functional materials for liquid breeder blanket concept for the future nuclear fusion reactor DEMO. Nevertless, the use of some shifter components will be needed to optimize some irradiation parameters.     

Inversion of neutron/gamma spectra from scintillator measurements

The Radiation Assessment Detector (RAD) on-board NASA’s Mars Science Laboratory (MSL) rover will measure charged particles as well as neutron and gamma radiation on the Martian surface. Neutral particles are an important contribution to this radiation environment. RAD measures them with a CsI (Tl) and a plastic scintillator, which are both surrounded by an anticoincidence. The incident neutron/gamma spectrum is obtained from the measurements using inversion methods which often fit a functional behavior, e.g., a power law, to the measured data applying the instrument response function and, e.g., a least-squares method. In situations where count rates are small, i.e., where the stochastic nature of the measurement is evident, maximum likelihood estimates with underlying Poissonian statistics improve the resulting spectra. We demonstrate the measurement and inversion of gamma/neutron spectra for a detector concept featuring one high-density scintillator and one high-proton-content scintillator. The applied inversion methods derive the original spectra without any strong assumptions of the functional behavior. Instrument response functions are obtained from Monte-Carlo simulations in matrix form with which the instrument response is treated as a set of linear equations. Using the response matrices we compare a constrained least-squares minimization, a chi-squared minimization and a maximum likelihood method with underlying Poissonian statistics. We make no assumptions about the incident particle spectrum and the methods intrinsically satisfy the constraint of non-negative counts. We analyzed neutron beam measurements made at the Physikalisch Technische Bundesanstalt (PTB) and inverted the measurement data for both neutron and gamma spectra. Monte-Carlo-generated measurements of the expected Martian neutron/gamma spectra were inverted as well, here the maximum likelihood method with underlying Poissonian statistics produces significantly better results.     

In vessel neutron instrumentation for sodium-cooled fast reactors: Type,lifetime and location

The in vessel instrumentation of sodium-cooled fast reactors must deliver measurements that are reliable and easy to interpret over several reactor cycles in order to fulfill the safety requirements. This paper compares, with respect to this requirement, three types of detectors that are widely used in neutron measurements: fission chambers, boron-lined proportional counters, self-powered neutron detectors. We use neutron spectra that are computed for preliminary design of sodium-cooled fast reactor in different representative locations: in diluting tubes within nuclear fuel assemblies, or in the lateral neutron protections. With an evolution code, we compute the expected signal for each type of detector, to assess whether its level is sufficient, and also its evolution over three operating cycles, to examine whether it is compatible with long term measurements. The conclusion is that fission chambers are the only type able to deliver an interpretable signal for a wide dynamic of reactor power and for three or more operating cycles. The two other types are shown to be inadequate.     

基于信赖域算法的中子解谱方法研究

提出了一种新的基于信赖域算法的中子能谱求解方法。使用O5S软件仿真了能量范围为0.25～16 MeV、能量间隔为0.25 MeV的BC501A液体闪烁体探测器响应函数。利用该响应函数仿真验证了入射中子能谱分别为单峰和多峰情况下算法的解谱效果，并通过D-T单能中子源、^{241Am-Be和252Cf连续中子源对解谱方法进行了实验验证。结果表明，使用信赖域算法求得的能谱与参考谱线具有较好的一致性，初步验证了解谱方法的有效性。}