Design of transition cores of RSG GAS (MPR-30) with higher loading silicide fuel

A procedure of designing transition cores to achieve the equilibrium silicide core of RSG GAS with higher fuel loading of 300 g U/fuel element (FE) (meat density of 3.55 g U/cm³) has been proposed. In the proposed procedure, the EOC excess reactivity of each transition core is minimized in order to satisfy the safety design limit of one-stuck-rod sub-criticality margin while keeping the maximum of radial power peaking factor below the allowable value. Under the design procedure, the initial fuel loadings are increased gradually in two steps, i.e. from 250 to 275 g U/FE followed by 275-300 g U/FE. The analysis results show that all transition cores can satisfy all design requirements and safety limits. We concluded that the obtained transition core design should be adopted into the future core conversion program of RSG GAS. The targeted silicide core can be achieved practically in at least 24 transition cores.     

核临界安全中的监督现场测量技术--源倍增方法的某些问题

给出了核临界安全中监督现场的测量技术——源倍增法的实验理论和实验方法。源倍增法实际测量的是有源次临界中子有效增殖系数k2而不是中子有效增殖系数Keff。在铀溶液核临界装置上进行了实验研究用源倍增法测量了次临界系统在外中子源作用下铀溶液不同液位的有源次临界中子有效增殖系数k2;用周期法测量了单位铀溶液位的反应性系数,然后用临界液位与次临界液位之差乘以单位铀溶液位的反应性系数,给出系统次临界液位时的反应性．由反应性给出传统观念上的中子有效增殖系数keff 。讨论了它们的差别及对核临界安全的影响。     

Uncertainty quantification of criticality in solid-moderated and -reflected cores at Kyoto University Critical Assembly

Uncertainty quantification is conducted for the criticality of excess reactivity and control rod worth obtained at the Kyoto University Critical Assembly (KUCA). By combining SRAC2006 and MARBLE code systems, the sensitivity coefficients of the cross sections for aluminum-27 (²⁷Al) comprising mainly of core components are large in the solid-moderated and -reflected cores (A cores) at KUCA. Also, the uncertainty is dominant in the uranium-235 isotope by the covariance data of JENDL-4.0, and a quantitative value is about 150 pcm induced by the JENDL-4.0 data library in the KUCA A cores, whereas the covariance data of ²⁷Al are not prepared in JENDL-4.0. Moreover, the effect of decreasing uncertainty is obtained by applying the cross-sectional adjustment method to the uncertainty analyses. From the results, a series of uncertainty quantifications is expected to clarify the uncertainty of sub-criticality in accelerator-driven system experiments with spallation neutrons in the KUCA A cores.     

Feasibility of Continuous Sub-criticality Monitoring using a Digital Reactivity Meter to Overcome the Criticality Accident

In the foregoing studies, it has been proved that the digital reactivity meter can be used for not only sub-criticality measurement but also continuous sub-criticality monitoring during criticality approach. Based on these studies, we investigated the applicability of a digital reactivity meter for continuous sub-criticality monitoring to intervene before a criticality accident occurs that is similar to the Tokai-mura accident in 1999. In our mock up numerical simulation, there are three calculation steps that are (1) reactivity transient calculation, (2) neutron transient calculation and (3) sub-criticality monitoring calculation. The reactivity transient was calculated using two-group diffusion nuclear constants and the neutron transient was calculated using the one-point reactor kinetics. In sub-criticality monitoring, three algorithms for reactivity evaluation were compared. We have investigated which algorithm is the most suitable to use in an actual system. In practice, we also advise use of some filtering algorithm to reduce the neutron transient fluctuation and a warning reactivity to know estimated sub-criticality as earlier as possible.     

Small high temperature gas-cooled reactors with innovative nuclear burning

Since the innovative concept of CANDLE (Constant Axial shape of Neutron Flux, nuclide densities and power shape During Life of Energy producing reactor) burning strategy was proposed, intensive research works have been continuously conducted to evaluate the feasibility and the performance of the burning strategy on both fast and thermal reactors. We learned that one potential application of the burning strategy for thermal reactors is for the High Temperature Gas-Cooled Reactors (HTGR) with prismatic/block-type fuel elements. Several characteristics of CANDLE burning strategy such as constant reactor characteristics during burn-up, no need for burn-up reactivity control mechanism, proportionality of core height with core lifetime, sub-criticality of fresh fuel elements, etc. enable us to design small sized HTGR with a high degree of safety, easiness of operation and maintenance, and long core lifetime which are required for introducing the reactors into remote areas or developing countries with limited infrastructures and resources. In the present work, we report our evaluation results on small sized block-type HTGR designs with CANDLE burning strategy and compared with other existing small HTGR designs including the ones with pebble fuel elements, under both uranium and thorium fuel cycles.     

Assessments of the kinetic and dynamic transient behavior of sub-critical systems (ADS) in comparison to critical reactor systems

The neutron kinetic and the reactor dynamic behavior of Accelerator Driven Systems (ADS) is significantly different from those of conventional power reactor systems currently in use for the production of power. It is the objective of this study to examine and to demonstrate the intrinsic differences of the kinetic and dynamic behavior of accelerator driven systems to typical plant transient initiators in comparison to the known, kinetic and dynamic behavior of critical thermal and fast reactor systems. It will be shown that in sub-critical assemblies, changes in reactivity or in the external neutron source strength lead to an asymptotic power level essentially described by the instantaneous power change (i.e. prompt jump). Shutdown of ADS operating at high levels of sub-criticality, (i.e. k_eff 0.99), without the support of reactivity control systems (such as control or safety rods), may be problematic in case the ability of cooling of the core should be impaired (i.e. loss of coolant flow). In addition, the dynamic behavior of sub-critical systems to typical plant transients such as protected or unprotected loss of flow (LOF) or heat sink (LOH) transients are not necessarily substantially different from the plant dynamic behavior of critical systems if the reactivity feedback coefficients of the ADS design are unfavorable. As expected, the state of sub-criticality and the temperature feedback coefficients, such as Doppler and coolant temperature coefficient, play dominant roles in determining the course and direction of plant transients. Should the combination of these safety coefficients be very unfavorable, not much additional margin in safety may be gained by making a critical system only sub-critical (i.e. k_eff0.95). A careful optimization procedure between the selected operating level of sub-criticality, the safety reactivity coefficients and the possible need for additional reactivity control systems seems, therefore, advisable during the early design phase of any ADS systems in order to assure a benign transient response of the particular ADS design under investigation to typical plant transient initiators.     

Prospects for the development of propulsion nuclear power systems

Questions concerning the development of propulsion nuclear power systems are examined. The priorities for promising systems for the naval and civilian fleets are determined-reliability and survivability, secretiveness, safety, technical-economic indicators, ecology. Each group of requirements is analyzed in detail. Special attention is focused on the direction of development of reactor cores for nuclear-powered propulsion systems. __________ Translated from Atomnaya énergiya, Vol. 103, No. 4, pp. 211–215, October, 2007.     

ADS启明星1#次临界反应堆缓发中子有效份额的测量

本文提出了利用改进的源倍增法测量次临界系统的绝对反应性与跳源法测量的相对反应性相比获得缓发中子有效份额β_eff的方法。用改进的源倍增法测量了ADS启明星1#次临界反应堆某次临界状态下的绝对反应性为-2.235×10^-3。在相同的次临界状态下,用跳源法测量了以β_eff为单位的反应性ρ/β_eff为-0.291 5 $,二者相比得到ADS启明星1#次临界反应堆的缓发中子有效份额为0.007667。利用MCNP建模计算的结果为0.007 783,两者在2%内符合。     

Safety performance comparation of MOX, nitride and metallic fuel based 25–100 MWe Pb–Bi cooled long life fast reactors without on-site refuelling

In this paper the safety performance of 25–100 MWe Pb–Bi cooled long life fast reactors based on three types of fuels: MOX, nitride and metal is compared and discussed. In the fourth generation NPP paradigm, especially for Pb–Bi cooled fast reactors, inherent safety capability is necessary against some standard accidents such as unprotected loss of flow (ULOF), unprotected rod run-out transient over power (UTOP), unprotected loss of heat sink (ULOHS). Selection of fuel type will have important impact on the overall system safety performance.

The results of safety analysis of long life Pb–Bi cooled fast reactors without on-site fuelling using nitride, MOX and metal fuel have been performed. The reactors show the inherent safety pattern with enough safety margins during ULOF and UTOP accidents. For MOX fuelled reactors, ULOF accident is more severe than UTOP accident while for nitride fuelled cores UTOP accident may push power much higher than that comparable MOX fuelled cores.     

HFETR控制棒钴吸收体安全性分析研究

高通量工程试验堆（HFETR）控制棒钴吸收体入堆至今已经20余年,本文对59Co的燃耗以及其燃耗对控制棒价值的影响进行了分析计算,结果表明,9#～14#控制棒的钴吸收体的平均燃耗和最大燃耗分别为4.02%和5.45%,4#和7#控制棒的钴吸收体的平均燃耗和最大燃耗分别为6.45%和10.38%;考虑钴吸收体燃耗的影响,9#～14#控制棒价值几乎不变,4#和7#控制棒价值下降0.15βeff（对于HFETR,1βeff=0.0071）;钴吸收体的燃耗使得堆芯次临界度下降0.16βeff,而反应堆的停堆棒位几乎不变,因此HFETR控制棒钴吸收体是安全的,且其燃耗对钴吸收体控制棒价值的影响较小,不影响反应堆的安全运行。      

Spatial neutronic coupling aspects in nuclear reactors

In this paper, an effort is made to gain insights about neutronic coupling and decoupling phenomena of nuclear reactors and its consequences on their safety and stability. The neutronic coupling and decoupling aspects are investigated using eigenvalue separation (EVS) methodology. Higher harmonic eigenvalues are calculated by the method of mode subtraction. The eigenvalue separation for a typical 1000 MWe PWR is calculated and its relations with reactor core shape and size and consequent effects on spatial stability are investigated. It is demonstrated quantitatively that it is necessary to optimize height to diameter (H/D) ratio to suppress the susceptibility to multimode oscillations and to enable ease in designing spatial control algorithm. Consequences of extreme H/D ratio are also addressed. Optimum shape of the reactor core is investigated and the evaluation of upper limit of about 1.3 for H/D ratio has been carried out for large PWR cores. Safety implications of neutronic loose coupling on departure from nucleate boiling ratio (DNBR) are also addressed.     

Risk assessment of nuclear power plants

The objective of the present work is to develop recommendations for controlling the safety of nuclear power plants on the basis of risk assessments and safety certification of nuclear power plants. The Kursk nuclear power plant is considered as an example of a nuclear power plant with an RBMK reactor. The concept of risk assessment of a nuclear power plant consists in constructing a set of scenarios of the appearance and development of possible accidents followed by an evaluation of the realization frequency and determination of the scales of the consequences of each one. The result of an analysis is an evaluation of a system of risk indicators in accordance with the requirements of the safety compliance certificate of the nuclear power plant as well as the development of recommendations for increasing plant safety. In risk assessment, the consequences are divided into categories of the seriousness of the damage, for which their probability is evaluated separately. The graphical interpretation of risk due to any dangerous object consists of frequency–consequences curves. Recommendations are developed on the basis of the results of risk analysis.     

Safety prediction in nuclear power

The practical operational safety of nuclear objects is of fundamental importance for assessing the future prospects under discussion and selecting a strategy for the development of nuclear power. It is shown that the methods currently being used for making safety predictions do not contain an analysis of the unavoidable errors and uncertainties of the models used or the initial and boundary conditions under which the physical processes that develop into serious accidents arise and develop. It is proposed that the method of quantile estimates of the uncertainties, which is free of the drawbacks of the Monte Carlo method and which increases the reliability of safety predictions in nuclear power, be used. __________ Translated from Atomnaya énergiya, Vol. 102, No. 2, pp. 80–86, February, 2007.     

Effects of dimensional change strain in nuclear graphite component stress analysis

Shrinkage and thermal stresses are induced into graphite components when they are irradiated in nuclear reactor cores. These stresses have to be taken into account in the reactor design and subsequent safety case assessments. This is usually done using graphite irradiation constitutive models programmed into a finite element code. The models use empirical data for the irradiation induced property and dimensional change, which are obtained from graphite material test reactor programmes. The dimensional change in nuclear graphite is one of the most important strains induced by the irradiation fluence. In this paper the effect of two different numerical methods to calculate the dimensional change strain is examined. Then the effect on the predicted stress using two different empirical models for dimensional change is studied. The solutions show that although the difference between two models is small, there are considerable differences in the stress profile.     

跳源法在启明星1#次临界装置中的应用

简要介绍了跳源法在启明星1#次临界装置上测量次临界度的原理、外源驱动的次临界中子学实验装置、堆芯布置及中子源驱动系统。主要研究了中子源在堆芯轴向中心位置、不同装载情况下的反应性变化,并给出不同的有效倍增系数keff。实验测量结果与理论计算结果进行了比较,结果符合较好。     

Hafnium clad fuels for fast spectrum BWRs

We show that by use of hafnium cladding, a fast neutron spectrum is achievable in the top of uprated BWRs. Monte Carlo calculations have been made for Hf clad inert matrix nitride and low fertile MOX fuels, with fuel segments located in the upper part of an uprated BWR, where the coolant void fraction exceeds 70%. The nitride fuel results in the hardest neutron spectrum, but the low fertile MOX fuel still yields fission probabilities for even neutron number nuclides similar to those of sodium cooled reactors. The inert matrix nitride fuel configuration yields high burning rates, permitting to stabilise TRU inventories with less than 50% BWR cores of the here suggested type in the power park. The core with low fertile MOX fuel is less efficient, but still a zero net producer of TRU. Fuel and coolant temperature feedbacks are affected by introduction of absorbing elements in the fuel, but remain within acceptable ranges for the low fertile MOX fuel. Although control rod worths are reduced, shutdown margins are sufficient to ensure sub-criticality in cold conditions. From a materials point of view, the behaviour of hafnium clad MOX fuel would be similar to zircalloy clad MOX fuel already used extensively in nuclear industry. Thus, if dynamic stability of the core can be ensured, the here proposed fuel may be considered as a low cost solution for transmutation of minor actinides on industrial scale.     

反应堆物理实验中的源倍增法研究

给出了反应堆物理实验中临界测量和次临界度测量通常所采用的源倍增方法研究。首先从有源的扩散理论出发,导出了与以前不同的源倍增方法的公式。源倍增方法测量的参数实际是次临界系统在外源作用下的有源次临界中子倍增因子ks,而不是在这之前的中子有效倍增因子keff,然后研究了实验装置的临界质量,研究了ks与外源位置和能谱的关系,证明了导出的源倍增方法的理论是正确的。该方法可像过去那样用于反应堆物理实验中的临界外推测量,但不能用于次临界度测量。解决了长期困扰人们有关源倍增方法测量的参数问题。最后讨论了ks和keff的差别和关系以及对临界外推测量和核临界安全的影响。     

Rossi-α测量方法的蒙特卡罗直接模拟

本文基于Geant4 toolkit开发了用于Rossi-α测量方法模拟的蒙特卡罗直接模拟程序,模拟计算了橡树岭实验室基准装置6个不同几何尺寸浓缩铀圆柱系统的瞬发中子衰减常数α,同时采用脉冲中子源方法模拟计算了α,二者结果一致。蒙特卡罗直接模拟计算结果与实验结果存在偏差,空隙是产生偏差的最主要原因,随着次临界度的加深,空隙的影响减小,计算和实际测量的α的相对偏差从19%变为0.19%。     

Numerical analysis of startup PWR with fractional neutron point kinetic equation

In this paper we present the numerical analysis of the neutron density behavior when the nuclear reactor power is increased during startup of a PWR. The fractional neutron point kinetic (FNPK) equation with one-group delayed neutron precursor and external neutron source was used for this analysis. It is considered that there is a relaxation time associated with a rapid variation in the neutron flux and this effect is considered with the FNPK which have a physical interpretation of the fractional order is related with the sub-diffusive process, i.e., non-Fickian effects from the neutron diffusion equation point of view. In order to study the relaxation time effects during start-up of a PWR, a numerical analysis with FNPK is carried out, which it is assumed that during the ith step of control rod withdrawal the way of reactivity insertion is step to step, where the neutron source strength was defined as a constant in terms of a known initial stable sub-criticality and the neutron signal from a steady state condition. The results of the FNPK were compared with the classical neutron point kinetics (CNPK), for different values of the anomalous relaxation time.     

Seismic safety of nuclear power plants in eastern Europe

This paper summarizes the work performed by the International Atomic Energy Agency in the areas of safety review and applied research in support of programmes for the assessment and enhancement of seismic safety in Eastern Europe and in particular, WWER type nuclear power plants during the past seven years. Three major topics are discussed; engineering safety review services in relation to external events, technical guidelines for the assessment and upgrading of WWER type nuclear power plants, and the Coordinated Research Programme on "Benchmark study for the seismic analysis and testing of WWER type nuclear power plants". These topics are summarized in a way to provide an overview of the past and present safety situation in selected WWER type plants which are all located in Eastern European countries. The main conclusion of this paper is that even though there is now a thorough understanding of the seismic safety issues in these operating nuclear power plants, the implementation of seismic upgrades to structures, systems and components are lagging behind, particularly for those cases in which re-evaluation indicated the necessity to strengthen the safety related structures or install new safety systems.