Development of an asymmetric multiple-position neutron source (AMPNS) method to monitor the criticality of a degraded reactor core

Tne analytical/experimental method has been developed to to monitor the subcritical reactivity and unfold the k_∞ distribution of a degraded reactor core. The method uses several fixed neutron detectors and a ²⁵²Cfneutron source placed sequentially in multiple positions in the core. Therefore, it is called the asymmetric multiple-position neutron source (AMPNS) method. The AMPNS method employs the nucleonic codes to analyze in two dimensions the neutron multiplication of a ²⁵²Cf neutron source. An optimization program, GPM, was utilized to unfold the k_∞ distribution of the degraded core, in which the desired performance measure minimizes the error between the calculated and the measured count rates of the degraded reactor core. The analytical/experimental approach is validated by performing experiments using the Penn. State Breazeale TRIGA reactor (PSBR). A significant result of this study has been to provide a means to plan the source and detector placements and assign core cells to the damaged TMI-2 core as well as to monitor the criticality during the recovery period.     

On the multiplication factor and reactivity definitions for subcritical reactor systems

An analysis is made on the merit of different functions adopted for weighting neutron processes in subcritical nuclear reactor systems, as it appears in expressions of relevant integral quantities, such as reactivity worths, prompt neutron lifetimes, etc. All weight functions may be shown to depend on some sort of explicit or implicit, real or fictitious, system control. Associated with the importance function relevant to the reactor power control, the multiplication factor k_sub and generalized reactivity values ρ_gen are defined. The difference (1-k_sub)/k_sub is shown to be the more appropriate index for generally representing the ADS subcriticality. However, in certain circumstances, when an accidental event is studied in which the criticality condition may be surpassed during the transient, it appears more appropriate to take into account the standard multiplication factor (k_eff) and the reactivity values to which the transient is associated and for the definition of which the standard adjoint flux is adopted.     

ADS反应堆物理研究中的稳态外源系统

给出了在ADS反应堆物理研究中所使用的稳态外中子源252Cf的有关问题。它包括了所使用的252Cf中子源的外形尺寸,包壳材料,焊接技术及试验所采用的标准;在ADS反应堆物理研究中传输252Cf中子源的“跑兔”系统;252Cf中子源储存容器设计及中子源位置的监督等问题。     

Prompt alpha and reactivity measurements on fast metal assemblies

This paper summarizes and reviews the methods of reactivity determination and measurement of the prompt-neutron decay, briefly describes the equipment requirements for such measurements for unmoderated metal assemblies, and presents experimental results to illustrate the methods. These assemblies include fast metal critical assemblies and fast pulsed reactors. The primary reactivity determination methods used have been: (1) stable reactor period measurements which are usually used near delayed criticality to obtain the reactivity or to calibrate the reactivity prior to burst initiation; (2) prompt reactor period measurements which are useful to determine the reactivity early in superprompt critical excursions; (3) inverse kinetics rod drop measurements which obtain the reactivity as a function of time after a rod or reactor component is removed from a delayed critical assembly, and (4) prompt neutron decay constant measurements from which the reactivity can be obtained if corrections are made for changes in the neutron lifetime. Inverse kinetics and decay constant measurements are usually used below delayed criticality, although decay constant measurements have been performed above delayed critical.The decay constant is usually obtained by the traditional pulsed-neutron method, using a pulsed neutron source such as a Cockcroft-Walton accelerator, or by the Rossi-α method. The recent use of ²⁵²Cf has resulted in some new techniques for determining the decay constant and reactivity; this method has some unique advantages over the traditional methods. The theory of these new measurements is reviewed, and some recent results are presented.     

Reactivity Measurements by Reactor Noise Analysis using Two-Detector Correlation Method

For a subcritical reactor, power spectral density measurement of neutron fluctuations can be used to evaluate the quantity α=(1 — ρ/β)/β/l, where ρ/β, i.e.the reactivity, can be determined using the quantity β/l measured at critical state.

In an actual experiment however, chamber noise is a hindrance, particularly at the highly sub-critical state. This makes large reactivity determination by reactor noise analysis difficult.

The correlation technique was used to eliminate the chamber noise component, thereby considerably extending the possible range of reactivity determination by reactor noise analysis.     

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.     

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.     

Subcritical multiplication factor and source efficiency in accelerator-driven system

The subcritical multiplication factor k_s   and the external neutron source efficiency φ^∗${\phi }^{\ast }$ are important parameters in the accelerator-driven system (ADS) performance assessment. The theoretical relation between k_s and the effective multiplication factor k_eff in a subcritical system is discussed in different cases of subcritical system. On the basis of the theoretical background, the dependence of k_s   and φ^∗${\phi }^{\ast }$ on subcriticality, source position, and energy is numerically investigated using a simple thermal subcritical model. For the sake of experimental evaluation of k_s   and φ^∗${\phi }^{\ast }$, the ADS experiments have been carried out in the subcritical systems combined with 14 MeV pulsed neutrons of the Kyoto University Critical Assembly (KUCA). The k_s   and φ^∗${\phi }^{\ast }$ parameters are successfully measured by utilizing the reaction rate distribution obtained by the optical fiber detectors in the subcritical system, within a relative difference of less than 7% and 12% for k_s   and φ^∗${\phi }^{\ast }$, respectively, between measured and calculated values for most studied cases.     

实验测量CFBR-Ⅱ堆启动Am-Be中子源的初始增殖倍数

由于CFBR-Ⅱ堆原启动中子源²⁵²Cf源半衰期较短,中子发射率已降低至无法满足使用要求,因此采用半衰期较长的Am-Be中子源替代。通过分别测量²⁵²Cf源与Am-Be源在裸源情形下的计数率以及处于活性区中心时引起的泄漏中子计数率,建立比例关系,借助于²⁵²Cf源对应的初始增殖倍数间接给出了Am-Be源对应的初始增殖倍数,为反应堆运行提供参数。     

Monitoring the reactivity of extremely subcritical reactors by means of reactivity meters,corrections being made to the analog of the source

Conclusions The advantage of the instrumental methods here described lies in the fact that no additional apparatus is required for determining subcriticality; it is sufficient simply to use an analog reactivity meter allowing the source component in the solving part to be varied. By using these methods we may determine the bias voltage (constituting the analog of a steady-state source in the reactor) in the subcritical state. If the intensity of the source does not alter very sharply over the period of measurement (5–10 min), the introduction of the source function into the instrument is an on-off operation, after which the reactivity meter is able to monitor any changes in K_eff continuously without first bringing the reactor into the critical state. This offers the possibility of making quite accurate measurements to ensure nuclear safety. In cases in which it is required to secure especially accurate and reliable results, these measurements may be repeated as from the critical state as part of a total monitoring operation. The safety of operations requiring passage into the critical state is also increased, since the passage into the critical state is monitored completely.Instrumental methods of measuring subcriticality are universal; they embrace not only critical assemblies but also any energy-stressed reactors in the subcritical state, and enable us to measure temperature effects, effects of poisoning, and so on, i. e., effects characteristic of energy-stressed reactors. In addition to this, these methods may be applied both in the presence and in the absence of a strong neutron background in the reactor.Translated from Atomnaya Énergiya, Vol. 41, No. 4, pp. 238–241, October, 1976.     

Estimation of the neutron chain-length distribution in subcritical systems using a point Monte Carlo code

A point Monte Carlo code has been developed that simulates the evolution of prompt fission chains in subcritical systems. From this simulation, a numerical estimate of the neutron chain-length distribution as a function of k_eff has been obtained for subcritical systems. We also discuss the importance of the detector efficiency in performing a successful neutron noise measurement in highly subcritical systems.     

A benchmark on the calculation of kinetic parameters based on reactivity effect experiments in the CROCUS reactor

Measurements in the CROCUS reactor at EPFL, Lausanne, are reported for the critical water level and the inverse reactor period for several different sets of delayed supercritical conditions. The experimental configurations were also calculated by four different calculation methods. For each of the supercritical configurations, the absolute reactivity value has been determined in two different ways, viz.: (i) through direct comparison of the multiplication factor obtained employing a given calculation method with the corresponding value for the critical case (calculated reactivity: ρ_calc); (ii) by application of the inhour equation using the kinetic parameters obtained for the critical configuration and the measured inverse reactor period (measured reactivity: ρ_meas). The calculated multiplication factors for the reference critical configuration, as well as ρ_calc for the supercritical cases, are found to be in good agreement. However, the values of ρ_meas produced by two of the applied calculation methods differ appreciably from the corresponding ρ_calc values, clearly indicating deficiencies in the kinetic parameters obtained from these methods.     

Small,long-life high temperature gas-cooled reactor free from prompt supercritical accidents by particle-type burnable poisons

A design concept for a high temperature gas-cooled reactor without the possibility of a prompt supercritical accident has been proposed by coupling the use of particle-type burnable poison (BP) and criticality control by the core temperature. The combinations of two different BPs, B₄C and Gd₂O₃ particles and B₄C and CdO particles, with the proper particle sizes and the appropriate volume ratio, showed excellent performance in controlling excess reactivity and flattening the reactivity swing. To maintain reactivity at a lower level than the prompt critical state, the reactor was designed to operate in a subcritical mode for a burnup period or for the whole operation cycle. Under subcritical operation during the partial burnup period, the core temperature had to be lowered by at least 164 K for the loading of B₄C + Gd₂O₃ particles and by at least 178 K for the B₄C + CdO particles, which in turn dropped the thermal efficiency from 48% to 42.26% and 41.77%, respectively. On the other hand, under full subcritical operation, a greater decrease of core temperature was required. Remarkable decreases in the core temperatures, approximately 347 K for the B₄C + Gd₂O₃ case and approximately 280 K for the B₄C + CdO case, resulted in the drop of thermal efficiency to only 35.9% and 38.2%, respectively. Therefore, the relative importance of the increase in passive safety and the decrease in thermal efficiency must be considered with regard to their importance in nuclear reactor design.     

基于CORCA的带固定中子源堆芯求解与共轭计算的软件实现

深度次临界状态下,传统源倍增法在核反应堆反应性测量上具有精度低的特点,为提高测量精度,本文对CORCA软件进行扩充,开发了具备固定源问题求解和带不连续因子中子价值求解功能的CORCAFIX软件,并采用对照程序和实堆数据对CORCA-FIX软件进行了计算验证。验证结果证实,CORCA-FIX在求解带固定源堆芯的深度次临界状态时有着较高的精度,输出的结果应用于实堆数据后获得了更好的次临界度测量结果,且满足工程应用中反应性测量的偏差准则。     

Absolute measurement of βeff and l on weapon-grade MOX fuel at the VENUS critical facility by means of the RAPJA technique

Measurements were performed at the VENUS critical facility for the validation of neutron codes in the case of a reactor core loaded with weapon-grade plutonium. Since the effective delayed neutron fraction β_eff and the mean neutron lifetime l are two important parameters in the characterisation of the dynamic behaviour of a reactor, an absolute measurement of these parameters is necessary to assure the correct validation of neutron transport codes. To meet this requirement, the RAPJA-technique was used. The RAPJA technique combines a Prompt Jump with a Rossi-Alpha measurement at the stationary subcritical level associated with the antireactivity level obtained after the Prompt Jump. The obtained results are well in accordance with calculated values for MOX-cores at VENUS. The measurements also show that uncertainties of 3 and 1.5% on respectively β_eff and l are achievable on a core loaded with weapon-grade plutonium.     

Critical neutron heating in the control drums of a dual purpose thermionic space reactor for power and propulsion

The critical neutron heating in the reflector control drums is investigated for a fast incore thermionic space craft reactor for power and nuclear propulsion. The reactor is fueled with uranium carbide (UC) and controlled with the help of rotating B₄C drums imbedded into the beryllium reflector. While the neutron heating in the drums would not require a cooling mechanism in the power phase, the heat generation during the thrust phase obliges cooling for a nuclear thermal thrust around F = 5000 N by a specific impulse of 670 s⁻¹ at an hydrogen exit temperature around 1900°K. With a beryllium reflector without extra cooling measures, thermal thrust must be kept F < 2500 N to relieve the thermal load in the reflector. On the other hand, a reflector made of BeO may withstand a thermal load for a nuclear thermal thrust of F = 5000 N. The neutronic analysis has been conducted in S₁₆-P₃ and S₈-P₃ approximation with the help of one- and two-dimensional neutron transport codes ANISN and DORT, respectively. A reactor control with boronated reflector drums (drum diameter = 14 cm) at the outer periphery of the radial reflector of 16 cm thickness would make possible reactivity changes of Δk_{eff = 13.55%}—amply sufficient for a fast reactor—without a significant distortion of the fission power profile during all phases of the space mission. Calculations are conducted for a reactor with a core radius of 22 cm and core height of 35 cm leading to power levels around 50 kW_el.     

Measurement of the effective delayed-neutron dose by the α-Rossi method in water-containing media

A modified method of determining the effective delayed-neutron dose is proposed. The method is based on a combination of determining the pseudoreactivity of the ²⁵²Cf source and measuring the α-Rossi time distribution. The advantage of the method proposed here is that there is no need to measure the absolute rate of fissioning of the nuclei in the medium. It is proposed that the latter be expressed in terms of the intensity of a known source of spontaneous fissions of ²⁵²Cf nuclei. The measurements are performed on an assembly with a hydrogen-containing moderator. The influence of the positions of the detectors (³He counters) on the measured effective delayed-neutron fraction is studied. __________ Translated from Atomnaya énergiya, Vol. 100, No. 5, pp. 393–399, May, 2006.     

Sectored Compact Space Reactor (SCoRe) concepts with a supplementary lunar regolith reflector

Design concepts of the Sectored Compact Space Reactor for Small power (SCoRe-S) have been developed for the avoidance of single-point failures in reactor cooling and energy conversion and a wide range of thermal powers. These modular, fast neutron spectrum, lithium cooled reactors with 16.0 cm thick BeO radial reflector are designed for at least +$2.00 hot-clean excess reactivity, and with a sufficient reactivity shutdown margin. They employ ¹⁵⁷GdN additives in the UN fuel and a 0.10 mm thick coating of ¹⁵⁷Gd₂O₃ on the outer surface of the reactor vessel to ensure that the bare reactors, when submerged in wet sand and flooded with seawater following a launch abort accident, remain at least ?$1.00 subcritical. In addition to identifying the smallest SCoRe-S concept that satisfies the design reactivity requirements, the benefit of using a lunar regolith as a supplementary reflector to decrease the thickness of the BeO radial reflector and hence, the launch mass of the SCoRe-S concepts for a lunar outpost is investigated. Calculations performed using MCNP5 confirmed that SCoRe-S₇ with a 16 cm thick BeO reflector is the smallest to satisfy the stated reactivity requirements. Results also show that a lunar regolith reflector alone is inadequate for this reactor to achieve a critical state at the beginning of life. However, when the regolith is used in conjunction with a BeO reflector of a reduced thickness, this reactor not only becomes critical, but also satisfies the reactivity design requirements at a significantly reduced launch mass. Using a supplementary reflector of regolith decreases the thickness of the BeO reflector for the SCoRe-S₇ from 16 cm to 8.0 cm, and to 5.7 cm for the SCoRe-S₁₁ of the largest core. The resulting decreases in the launch mass of the SCoRe-S concepts are ～34% or 150–200 kg.     

Neutronic calculation to the TRIGA Ipr-R1 reactor using the WIMSD4 and CITATION codes

The WIMSD4 and CITATION codes are used to calculate neutronic parameters of a TRIGA reactor. The results are compared with experimental values. Five configurations are analysed and the excess reactivity worth, the fuel temperature reactivity coefficient, the control reactivity worth, safety and regulation rod of the TRIGA IPR–R1 reactor are calculated. The idea is to obtain the systematic error for k_∞ for this methodology comparing the calculated and the experimental results.