Effects of high density dispersion fuel loading on the kinetic parameters of a low enriched uranium fueled material test research reactor

The effects of using high density low enriched uranium on the neutronic parameters of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density LEU fuels currently being developed under the RERTR program. Since the alloying elements have different cross-sections affecting the reactor in different ways, therefore fuels U–Mo (9 w/o) which contain the same elements in same ratio were selected for analysis. Simulations were carried out to calculate core excess reactivity, neutron flux spectrum, prompt neutron generation time, effective delayed neutron fraction and feedback coefficients including Doppler feedback coefficient, and reactivity coefficients for change of water density and temperature. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the excess reactivity at the beginning of life does not increase as the uranium density of fuel. Both the prompt neutron generation time and the effective delayed neutron fraction decrease as the uranium density increases. The absolute value of Doppler feedback coefficient increases while the absolute values of reactivity coefficients for change of water density and temperature decrease.     

Prospects of using different clad materials in a material test research reactor – Part 2 – The reactivity feedback coefficients

The reactivity feedback coefficients of a material test research reactor using stainless steel-316 and zircaloy-4 as clad were calculated. For this purpose, the aluminum clad of an MTR was replaced with stainless steel-316 and zircaloy-4. Calculations were carried out to find the fuel temperature reactivity feedback coefficient, clad temperature reactivity feedback coefficient, moderator temperature reactivity feedback coefficient and moderator density reactivity feedback coefficient. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the average values of fuel temperature reactivity feedback coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient from 38 °C to 50 °C, at the beginning of life, were maximum in magnitude for stainless steel-316 cladded fuel, followed by aluminum and least for the zircaloy-4 cladded fuel. The fuel temperature feedback coefficient increased in magnitude by 47.37% for stainless steel-316 and decreased by 4.72% for zircaloy-4 clad. The moderator temperature feedback coefficient increased in magnitude by 60.41% for stainless steel-316 and decreased by 3.03% for zircaloy-4 clad, while the moderator density feedback coefficient showed an increase in magnitude of 59.18% for stainless steel-316 and a decrease of 7.63% for zircaloy-4 clad. Zircaloy-4 gave a positive value for clad temperature feedback coefficient, while the others two did not have any clad temperature feedback coefficient.     

Measurements of the isothermal, power and temperature reactivity coefficients of the IPR-R1 TRIGA reactor

The aim of this paper is to present the experimental results of the isothermal, power and temperature coefficients of reactivity of the IPR-R1 TRIGA reactor at the Nuclear Technology Development Center - CDTN in Brazil. The measured isothermal reactivity coefficient, in the temperature range measured, was −0.5 ¢/°C, and the reactivity measurements were performed at 10 W to eliminate nuclear heating. The reactor forced cooling system was turned off during the measurements. When the reactor is at zero power there is no sensible heat being released in the fuel, and the entire reactor core can be characterized by a single temperature. The power coefficient of reactivity obtained was approximately −0.63 ¢/kW, and the temperature reactivity coefficient of the reactor was −0.8 ¢/°C. It was noted that the rise in the coolant temperature has contributed only with a small fraction to the observed negative effect of the reactivity. The power defect, which is the change in reactivity taking place between zero power and full power (250 kW), was 1.6 $. Because of the prompt negative temperature coefficient, a significant amount of reactivity is needed to overcome temperature and allow the reactor to operate at the higher power levels in steady state.     

A stochastically perturbed nonlinear point reactor model

We present an exact analysis of the fluctuations of neutron density in a stochastically perturbed nonlinear point reactor model in the absence of delayed neutrons. The reactivity and feedback coefficients are assumed to have white noise Gaussian component. The time development of probability distribution is obtained from the Fokker-Planck equation. It is found that for small density of neutron population the stationary probability distribution is more sensitive to the random change in the reactivity than in the feedback coefficient, whereas for large neutron density the probability distribution is more sensitive to the random change in the feedback coefficient than in the reactivity.     

Modeling of void, fast power and graphite temperature reactivity coefficients measurements for the validation of RELAP5-3D RBMK-1500 reactor model

This paper deals with the modeling of RBMK-1500 specific transients taking place at Ignalina NPP: measurements of void and fast power reactivity coefficients, as well as change of graphite cooling conditions transient. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and based on the obtained experimental results the actual values of these reactivity coefficients are determined. Graphite temperature reactivity coefficient at the plant is determined by changing graphite cooling conditions in the reactor cavity. This type of transient is unique and important from the point of view of model validation for the gap between fuel channel and the graphite bricks. The measurement results, obtained during this transient, enabled to determine the thermal conductivity coefficient for this gap and to validate the graphite temperature reactivity feedback model. The performed validation of RELAP5-3D model of Ignalina NPP RBMK-1500 reactor allowed to improve the model, which in the future would be used for the safety substantiation calculations of RBMK-1500 reactors.     

Reactivity feedback coefficients of a material test research reactor fueled with high-density U3Si2 dispersion fuels

The reactivity feedback coefficients of a material test research reactor fueled with high-density U₃Si₂ dispersion fuels were calculated. For this purpose, the low-density LEU fuel of an MTR was replaced with high-density U₃Si₂ LEU fuels currently being developed under the RERTR program. Calculations were carried out to find the fuel temperature reactivity coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the average values of fuel temperature reactivity feedback coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient from 20 °C to 100 °C, at the beginning of life, followed the relationships (in units of Δk/k × 10⁻⁵ K⁻¹) −2.116 − 0.118 ρ_U, 0.713 − 37.309/ρ_U and −12.765 − 34.309/ρ_U, respectively for 4.0 ≤ ρ_U (g/cm³) ≤ 6.0.     

Temperature and burnup reactivities and operational lifetime for the submersion-subcritical, safe space (S4) reactor

The He–Xe gas-cooled, S4 reactor has a sectored, Mo–14%Re solid core for avoidance of single point failures in reactor cooling and Closed Brayton Cycle (CBC) energy conversion. The reactor core is loaded with UN fuel and each of its three sectors is thermal-hydraulically coupled to a separate CBC loop and radiator panels. The solid core minimizes voids, and the BeO reflectors are designed to easily disassemble upon impact, ensuring that the bare S4 reactor is sufficiently subcriticial when submerged in wet sand or seawater and flooded with seawater, following a launch abort accident. Spectral shift absorber (SSA) additives in the core and thin SSA coatings on the outer surface of the core can also be used to ensure subcriticality in such an accident. This paper investigates the effects of various SSAs (Re, Ir, Eu-151, B-10 and Gd-155) on the temperature and burnup reactivity coefficients and the operating lifetime of the S4 reactor at a steady thermal power of 550 kW. The calculations of the burnup, reactivity feedback coefficient used a mixture of the top 10 light and top 10 heavy fission products plus Sm-149 and are performed for isothermal reactor core and reflector temperatures of 1200 and 900 K. In this fast spectrum space reactor, SSAs markedly increase fuel enrichment and decrease the burnup reactivity coefficient, but only slightly decrease the temperature, reactivity feedback coefficient. With no SSAs, the UN fuel enrichment is lowest (58.5 wt.%), the temperature and burnup reactivity coefficients are the highest (−0.2709 ¢/K and −1.3470 $/at.%), and the estimated operating lifetime is the shortest (7.6 years). The temperature and burnup reactivity coefficients decrease to −0.2649 ¢/K and −1.0230 $/at.%, and the operating lifetime increases to 8.3 years when rhenium additives are used. With europium-151 and gadolinium-155 additions, fuel enrichment (91.5 and 94 wt.%) and operating lifetime (9.9 and 9.8 years) are the highest and both the temperature reactivity feedback coefficient (−0.2382 and −0.2447 ¢/K) and the burnup reactivity coefficient (−0.9073 and −0.8502 $/at.%) are the lowest.     

Measurements of the isothermal,power and temperature reactivity coefficients of the IPR-R1 TRIGA reactor

The aim of this paper is to present the experimental results of the isothermal, power and temperature coefficients of reactivity of the IPR-R1 TRIGA reactor at the Nuclear Technology Development Center – CDTN in Brazil. The measured isothermal reactivity coefficient, in the temperature range measured, was −0.5 ¢/°C, and the reactivity measurements were performed at 10 W to eliminate nuclear heating. The reactor forced cooling system was turned off during the measurements. When the reactor is at zero power there is no sensible heat being released in the fuel, and the entire reactor core can be characterized by a single temperature. The power coefficient of reactivity obtained was approximately −0.63 ¢/kW, and the temperature reactivity coefficient of the reactor was −0.8 ¢/°C. It was noted that the rise in the coolant temperature has contributed only with a small fraction to the observed negative effect of the reactivity. The power defect, which is the change in reactivity taking place between zero power and full power (250 kW), was 1.6 $. Because of the prompt negative temperature coefficient, a significant amount of reactivity is needed to overcome temperature and allow the reactor to operate at the higher power levels in steady state.     

Measurement and analysis of feedback reactivity in the Monju restart core

The purpose of this study is to develop a feedback reactivity measurement technique in the Japanese prototype fast breeder reactor Monju and to validate calculation methodology to forecast the nuclear feedback phenomena. A feedback reactivity measurement technique has been developed based on a reactivity model featuring components that depend on the reactivity coefficients, denoted as reactor power (K_R) and reactor vessel inlet temperature (K_in). This technique can precisely measure the two reactivity components simultaneously and was applied to the feedback reactivity experiment conducted in the Monju system start-up test in May 2010. A thorough evaluation considering all possible biases and uncertainties demonstrated that the reactivity coefficients can be evaluated with a measurement uncertainty smaller than 3%. The evaluated reactivity coefficients were simulated considering the temperature distribution in the core. The calculated and measured values of K_R agreed within 1%, and the value of K_in was consistent with that reported in a previous isothermal temperature coefficient experiment. The measured and calculated fuel subassembly outlet temperatures also agreed well within 0.2°C, which supports the validity of the temperature calculation.     

Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity (ρ_ex), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 × 10³Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.     

Feedback reactivity coefficients and their coupling

Coupled reactivity feedback coefficients which accounts for variation in fuel temperature and moderator void simultaneously, have been determined for swimming pool type research reactor namely Pakistan Research Reactor PARR-1. The state of art is core criticality calculations, employing lattice cell code WIMS-D/4 and application of Taylor series expansion for core reactivity up to third order, involving two variables, i.e. fuel temperature and coolant void. The spectral effects in one region due to change of parameter in other region have also been studied. When spectral changes in moderator region due to 20 K change in fuel temperature have been incorporated in the calculation of fuel temperature coefficient, the results seems to be improved by 4.12%. Further, the results of void coefficient of reactivity show the improvement of 0.1% when the spectral effect in fuel region due to 5% change in void in moderator region is taken into account. These differences seem to be an improvement in the results, as physically any change in one region is accompanied by change in the other region.     

Experimental study of neutronic parameters in Tehran research reactor mixed-core

In this work, general characteristics of a typical mixed core, including HEU & LEU fuel is studied. The study is performed in the Tehran research reactor (TRR). In this study the neutronic parameters, reactivity feedback coefficients and kinetic parameters are investigated. The reference core designated for such study is the equilibrium core (No. 61) with an average bun-up of 27% & 36% for SFE's & CFE's, respectively. The MTR_PC package is used for neutronic analysis. In this research, experimental and computational results for the reference and mixed core are compared. Meantime, the obtained values for neutronic parameters are mostly below the adopted safety criteria and they are in good agreement with the experimental results. However β_eff and ℓ_p are a little bit higher in the mixed core with respect to the reference core, but in practice, these small changes will not cause substantial impacts on the dynamic behaviour of the reactor core. The absolute values of the fuel temperature, moderator density and void coefficients of reactivity, are less in the mixed core and only the moderator temperature coefficient is higher. The calculated values of power defect, based on the reactivity coefficients; in both core configurations are in good agreement with the experimental values.     

Reactivity feedbacks of a material test research reactor fueled with various low enriched uranium dispersion fuels

The reactivity feedbacks of a material test research reactor using various low enriched uranium fuels, having same uranium density were calculated. For this purpose, the original aluminide fuel (UAl_x–Al) containing 4.40 gU/cm³ of an MTR was replaced with silicide (U₃Si–Al and U₃Si₂–Al) and oxide (U₃O₈–Al) dispersion fuels having the same uranium density as of the original fuel. Calculations were carried out to find the fuel temperature reactivity feedback, moderator temperature reactivity feedback, moderator density reactivity feedback and moderator void reactivity feedback. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that the magnitudes all the respective reactivity feedbacks from 38 °C to 50 °C and 100 °C, at the beginning of life, of all the fuels were very close to each other. The fuel temperature reactivity feedback of the U₃O₈–Al was about 2% more than the original UAl_x–Al fuel. The magnitudes of the moderator temperature, moderator density and moderator void reactivity feedbacks of all the fuels, showed very minor variations from the original aluminide fuel.     

A power perturbation technique for the measurement of fuel-to-coolant heat transfer coefficients in fast reactors

This technique provides a method of obtaining average fuel to coolant heat transfer coefficients for individual fuel subassemblies in fast reactors. A series of experiments on the UK prototype fast reactor (PFR) over the period 1977–1979 have demonstrated that the technique is simple, requires no special instrumentation other than thermocouples to monitor coolant outlet temperatures, and the measurement can be made during normal reactor operation. Thus it is possible to determine how heat transfer coefficients change with operating conditions and with the degree of burn-up in the fuel.The analysis of a single experiment is presented to illustrate the technique. This was conducted at a single reduced power level of 200 thermal megawatts for two different primary coolant flow rates, both steady fractions of the maximum (0.88 and 0.47). Cyclic and single-step perturbations of about 10% amplitude were impressed on the steady power and the delayed coolant temperature response at subassembly outlets was monitored. Burn-ups in the subassemblies ranged between 1.0% and 4.7%. From the measured delays at the two flows it was possible to determine the fuel time-constant and hence the fuel-to-coolant heat transfer coefficient. It was also shown that a simple, lumped-element, heat transfer model can be used to obtain sufficiently accurate estimates from measurements at just one coolant flow.Fuel surface-to-coolant thermal conductances (i.e. gap conductances) were subsequently derived from the heat transfer coefficients. These ranged between 2.4 kW m⁻² K⁻¹ and 3.3 kW m⁻² K⁻¹ with the smaller conductances being obtained for those fuel elements with the larger degree of burn-up. These values are lower than expected but consistent with a higher than expected value for the negative power coefficient of reactivity feedback which has been observed at reduced power.     

Estimating Temperature Reactivity Coefficients by Experimental Procedures Combined with Isothermal Temperature Coefficient Measurements and Dynamic Identification

A method to evaluate the moderator temperature coefficient (MTC) and the Doppler coefficient through experimental procedures performed during reactor physics tests of PWR power plants is proposed. This method combines isothermal temperature coefficient (ITC) measurement experiments and reactor power transient experiments at low power conditions for dynamic identification. In the dynamic identification, either one of temperature coefficients can be determined in such a way that frequency response characteristics of the reactivity change observed by a digital reactivity meter is reproduced from measured data of neutron count rate and the average coolant temperature. The other unknown coefficient can also be determined by subtracting the coefficient obtained from the dynamic identification from ITC. As the proposed method can directly estimate the Doppler coefficient, the applicability of the conventional core design codes to predict the Doppler coefficient can be verified for new types of fuels such as mixed oxide fuels.

The digital simulation study was carried out to show the feasibility of the proposed method. The numerical analysis showed that the MTC and the Doppler coefficient can be estimated accurately and even if there are uncertainties in the parameters of the reactor kinetics model, the accuracies of the estimated values are not seriously impaired.     

Power Noise Spectra of a Water Reactor in Low Frequency Region

The power noise measurements were carried out on the Kyoto University Reactor at various reactor power levels under natural convection for two kinds of core configuration with different temperature coefficients of reactivity. Analysis of the results revealed strong noise in the low frequency region at higher power levels, even with a core configuration of essentially zero temperature coefficient of total reactivity.

A conventional theoretical model assuming a white noise input in reactivity and a reactor transfer function with temperature-reactivity feedbacks was adopted for comparison with the observed data, but it resulted in considerable disagreement at higher power levels.

As a result, it is concluded that there must exist other power effects not yet clarified in the reactivity feedbacks or in the noise inputs to the reactor, which are responsible for the experimental noise spectrum distortion in the low frequency region.     

Transient tests on blower trip and rod removal at the HTR-10

Safety demonstration tests on the 10 MW high temperature gas-cooled reactor test module (HTR-10) were conducted to verify the inherent safety features of MHTGRs and to obtain the core and primary cooling system transient data for validation of safety analysis codes.Two simulated anticipated transients without scram (ATWS) tests, lose of forced cooling by trip of the helium blower and reactivity insertion via control rod withdrawal were performed. This paper describes the tests with detailed test method, condition and results.Calculated results show that the strongly negative temperature coefficient causes reactor power to closely follow heat removal levels. Maximum fuel temperature changes are limited by the large core heat capacity to below 1230 °C during two tests.The test of tripping the helium circulator ATWS test was conducted on October 15, 2003. Although none of 10 control rods was moved, the reactor power immediately decreased due to the negative temperature coefficient. After about 50 min, the reactor became criticality again. Finally, the reactor power went to a stable level with about 200 kW.The test of reactivity insertion ATWS test was conducted two times. Following the control rod withdrawal, the reactor power increased rapidly, the maximum power level reached to 5037 and 7230 kW from the initial power of 3000 kW in accordance with reactivity insertion of $ 0.136 and 0.689, respectively. After the reactivity introduced was compensated by means of the strong negative reactivity feedback effect, the reactor went to subcritical and the power decreased.     

Liquid salt cooled flexible conversion ratio fast reactor: Neutronic design

This paper presents the neutronic design of a liquid salt cooled fast reactor with flexible conversion ratio. The main objective of the design is to accommodate interchangeably within the same reactor core alternative transuranic actinides management strategies ranging from pure burning to self-sustainable breeding. Two, the most limiting, core design options with unity and zero conversion ratios are described. Ternary, NaCl-KCl-MgCl₂ salt was chosen as a coolant after a rigorous screening process, due to a combination of favourable neutronic and heat transport properties. Large positive coolant temperature reactivity coefficient was identified as the most significant design challenge. A wide range of strategies aiming at the reduction of the coolant temperature coefficient to assure self-controllability of the core in the most limiting unprotected accidents were explored. However, none of the strategies resulted in sufficient reduction of the coolant temperature coefficient without significantly compromising the core performance characteristics such as power density or cycle length. Therefore, reactivity control devices known as lithium thermal expansion modules were employed instead. This allowed achieving all the design goals for both zero and unity conversion ratio cores. The neutronic feasibility of both designs was demonstrated through calculation of reactivity control and fuel loading requirements, fluence limits, power peaking factors, and reactivity feedback coefficients.     

快堆三维瞬态反应性反馈计算研究

在处理快堆时空动力学计算的反应性反馈问题时，提出了一种反应性直接反馈的数学模型。结合快堆的反应性反馈机制，在快堆中子学软件NAS的基础上，给出一种在时空动力学计算中截面反馈与反应性直接反馈相结合的反馈模式。同时，将快堆并群系统加入到程序中，实现了在线并群。对中国实验快堆(CEFR)等温温升过程进行模拟，通过计算结果与CEFR温度反应性系数实验测量结果的对比，证明了本模型和程序的正确性。     

核释热瞬态特性模拟初探

在利用热工水力台架进行反应堆热工水力研究时，一般采用电加热代替核释热。由于反应堆本身存在温度效应、空泡效应等内部反馈，因此用电加热代替核释热时，还应以模拟这种内部反馈。本文分析了各种内部反馈的不同影响，提出了模拟的重点，并在此基础上提出了一个模拟内部反部反馈的方法。