An analysis of rewetting of a nuclear fuel rod in water reactor emergency core cooling

An exact solution of the quasi-steady two-dimensional conduction equation for the rewetting of a nuclear fuel rod in water reactor emergency core cooling is obtained for a cylindrical rod geometry. The analysis adopts the conventional model of two heat transfer regions: zero heat transfer coefficient over the dry surface and a constant heat transfer coefficient over the wetted surface. Both the wet front velocity and the temperature distribution in the rod are computed. The present solution is valid over the whole range of Biot number.     

Cladding tube deformation and core emergency cooling in a loss of coolant accident of a pressurized water reactor

The paper summarizes the dominant effects which finally ensure the core coolability of a pressurized water reactor in a loss-of-coolant accident (LOCA).The main results are summarized as follows:

• — The cooling effect of the two-phase mixture which is intensified during reflooding increases temperature differences on the cladding tube circumference and thus limits the mean circumferential burst strains to values of about 50%.

• — An unidirected flow through the fuel rod bundle during the refill and reflooding phases causes maximum cooling channel blockage of about 70%.

• — The coolability of deformed fuel elements can be maintained up to flow blockages of about 90%.

All effects investigated indicate that in a LOCA no impairment of core coolability and public safety has to be expected.     

The design and experimental validation of an emergency core cooling system for a pool type research reactor

This paper presents the design of the Emergency Core Cooling System (ECCS) for the IEA-R1m pool type research reactor. This system with passive features, uses sprays installed above the core. The experimental program performed to define system parameters and to demonstrate to the licensing authorities, that the fuel elements limiting temperature is not exceeded, is also presented. Flow distribution experiments using a core mock-up in full-scale were performed to define the spray header geometry and spray nozzles specifications as well as the system total flow rate. Another set of experiments using electrically heated plates simulating heat fluxes corresponding to the decay heat curve after full power operation at 5 MW was conducted to measure the temperature distribution at the most critical position. The observed water flow pattern through the plates has a very peculiar behavior resulting in a temperature distribution which was modelled by a 2D energy equation numerical solution. In all tested conditions the measured temperatures were shown to be below the limiting value.     

应急堆芯冷却系统水锤隐患的分析和防范

应急堆芯冷却系统(ECC系统)是电站的专设安全系统,如果在注射过程中发生水锤,可能造成管道破裂,将严重影响到ECC系统的功能完整性,影响到堆芯的安全。根据ECC系统的功能和特点以及水锤产生的原理,对秦山第三核电厂ECC系统中存在的水锤隐患进行了分析,并提出了消除水锤隐患的建议修改方案,以防范注射过程中发生水锤。     

The effect of thermal diffusion from fuel pellets on rewetting of overheated water reactor pins

This paper presents the results of a finite difference solution of a conduction equation for the rewetting of a hot tube containing a filler material. The results show that the effect of a filler is always to reduce the rewetting velocity compared with an empty tube and reasonable agreement is obtained with previous experimental work. The effects of a gas gap on the rewetting of a UO₂-filled Zircaloy tube are discussed. A simple physical model is also presented which shows that the dominant parameter in determining the effect of a filler is (kpc)^1/2. It is suggested that previous theories for rewetting rate derived for empty tubes can be modified to include the effects of a filler by the use of a conduction correction term.     

Analytical evaluation on dynamical response characteristics of reduced-moderation water reactor with tight-lattice core under natural circulation core cooling

The time-domain analyses with TRAC-BF1 code were performed for clarifying the dynamical response characteristics of the reduced-moderation water reactor (RMWR) with tight-lattice core configuration. The response characteristics were evaluated based on the step response basically utilized for dynamical system evaluation. As for the most fundamental dynamical characteristics, the channel flow response characteristics of single fuel assembly were evaluated. In the evaluation, the appropriate single-phase pressure drop setting at the inlet orifice was determined in terms of response stability from the design viewpoint. In addition, from the investigation on the relation of the response and transit time of coolant, it is confirmed that the channel flow response of RMWR is dominated by the transit time of vapor phase resulting from a high void fraction operation condition. As for a natural circulation flow response, it is clarified that the response is strongly influenced by the effect of two-phase pressure loss owing to a high void fraction condition. The reactor power response with reactivity feedback shows quite stable response characteristics on account of the small absolute value of void reactivity coefficient.     

压水堆堆芯损伤评价系统研究与开发

堆芯损伤评价是反应堆事故后应急评价的重要组成部分。本文在国外文献基础上,结合国内的运行经验,对压水堆堆芯损伤评价进行研究,并开发了相应的软件程序。堆芯损伤评价包括基于堆芯裸露时间、在线监测仪表读数和取样分析数据三种方法。考虑应急实时要求、电厂实际情况与国际经验,本文采用了基于在线监测仪表读数的评价方法,该方法主要是基于堆芯热电偶读数与安全壳辐射监测仪表读数进行评价,其他监测仪表读数进行辅助合理性证实。     

Effect of Pu-rich agglomerates in MOX fuel on reactivity analysis of light water reactor MOX core physics experiments

The effect of Pu-rich agglomerates in U-Pu mixed oxide (MOX) on the reactivity analysis of light water reactor MOX core physics experiments was studied with the continuous-energy Monte Carlo calculation code MVP II. First, the following three different models were compared in the analysis of a representative unit cell of a MOX core tested at the KRITZ reactor: a Lattice model where Pu-rich agglomerates were assumed to exist in a fixed pitch, a statistical geometry (STG) model of MVP II, and a Random model where the random distribution of Pu-rich agglomerates was directly modeled. Since the three models gave comparable results, the STG model was used in parametric calculations to systematically understand the reactivity effect depending on the characteristics of Pu-rich agglomerates. In addition, the selected unit cells composing the MOX cores and one representing MOX core tested at the EOLE criticality facility were analyzed with the measured characteristics of Pu-rich agglomerates in MOX fuel. Consequently, the reactivity differences between the calculations assuming the homogeneous Pu distributions and those considering Pu-rich agglomerates were less than 0.0005 Δk/k/k', indicating that the effect of Pu-rich agglomerates was small on the reactivity analysis of the MOX cores tested in the EOLE facility.     

The reactor physics characteristics of a transuranic mixed oxide fuel in a heavy water moderated reactor

The reprocessing actinide materials extracted from spent fuel for use in mixed oxide fuels is a key component in maximizing the spent fuel repository utility. While fast spectrum reactor technologies are being considered in order to close the fuel cycle, and transmute these actinides, there is potential to utilize existing pressurized heavy water reactors such as the CANDU^®1 design to meet these goals. The use of current thermal reactors as an intermediary step which can burn actinide based fuels can significantly reduce the fast reactor infrastructure needed. This paper examines the features of actinide mixed oxide fuel, TRUMOX, in a typical CANDU nuclear reactor. The actinide concentrations used were based on extraction from 30 year cooled spent fuel and mixed with natural uranium in 4.75% actinide MOX fuel. The WIMS-AECL model of the fuel lattice was created and the two neutron group properties were transferred to RFSP in order to create a 3 dimensional time average full core model. The model was created with typical CANDU limits on bundle and channel powers and a burnup target of 45 MWd/kgHE. The TRUMOX fuel design achieved its goals and performed well under normal operations simulations. This effort demonstrated the feasibility of using the current fleet of CANDU reactors as an intermediary step in burning reprocessed spent fuel and reducing actinide burdens within the end repository. The recycling, reprocessing and reuse of spent fuels produces a much more sustainable and efficient nuclear fuel cycle using existing and proven reactor technologies.     

Core cooling in pressurized-water reactor during water injection

In this paper,the reactor core cooling and its melt progression terminating is evaluated,and the initiation criterion for reactor cavity flooding during water injection is determined.The core cooling in pressurized-water reactor of severe accident is simulated with the thermal hydraulic and severe accident code of SCDAP/RELAP5.The results show that the core melt progression is terminated by water injection,before the core debris has formed at bottom of core,and the initiation of reactor cavity flooding is indicated by the core exit temperature.     

The effect of precursory cooling on rewetting of a slab

The Dua and Tien (1976) model for the rewetting of a slab with precursory cooling is solved exactly by separation of variables. The solution for the rewetting velocity is found to agree very well with a Wiener Hopf technique solution to this model by the author. Rewetting rates predicted by the approximate solution of Dua and Tien are found to agree with the present solution for small Peclet numbers, while underpredicting them for large Peclet numbers. Theoretical quench front velocities compare favorably with experimental data for copper quenched by liquid nitrogen. Precursory cooling is shown to be able to greatly increase the rewetting velocity, in particular for cases of high flow rates, while neglecting it in modelling may result in much too low quench velocities, as compared to experimental measurements.     

Design evaluation of emergency core cooling systems using Axiomatic Design

In designing nuclear power plants (NPPs), the evaluation of safety is one of the important issues. As a measure for evaluating safety, this paper proposes a methodology to examine the design process of emergency core cooling systems (ECCSs) in NPPs using Axiomatic Design (AD). This is particularly important for identifying vulnerabilities and creating solutions. Korean Advanced Power Reactor 1400 MWe (APR1400) adopted the ECCS, which was improved to meet the stronger safety regulations than that of the current Optimized Power Reactor 1000 MWe (OPR1000). To improve the performance and safety of the ECCS, the various design strategies such as independency or redundancy were implemented, and their effectiveness was confirmed by calculating core damage frequency. We suggest an alternative viewpoint of evaluating the deployment of design strategies in terms of AD methodology. AD suggests two design principles and the visualization tools for organizing design process. The important benefit of AD is that it is capable of providing suitable priorities for deploying design strategies. The reverse engineering driven by AD has been able to show that the design process of the ECCS of APR1400 was improved in comparison to that of OPR1000 from the viewpoint of the coordination of design strategies.     

Monte Carlo simulation of core physics parameters of the Syrian MNSR reactor

A 3-D neutronic model for the Syrian Miniature Neutron Source Reactor (MNSR) was developed earlier to conduct the reactor neutronic analysis using the MCNP-4C code. The continuous energy neutron cross sections were evaluated from the ENDF/B-VI library. This model is used in this paper to calculate the following reactor core physics parameters: the clean cold core excess reactivity, calibration of the control rod and calculation its shut down margin, calibration of the top beryllium shim plate reflector, the axial neutron flux distributions in the inner and outer irradiation positions and calculations of the prompt neutron life time (_lp$l_p$) and the effective delayed neutron fraction (_βeff${\beta }_{\mathit{eff}}$). Good agreements are noticed between the calculated and the measured results. These agreements indicate that the established model is an accurate representation of Syrian MNSR core and will be used for other calculations in the future.     

多群核数据不确定性对堆芯物理计算的影响

核数据不确定性是造成反应堆物理计算结果不确定性的重要因素之一。基于所需抽样核数据的协方差矩阵开发了随机抽样模块(Stochastic Sampling,SAMP),在此基础上利用SCALE(Standardized Computer Analyses for Licensing Evaluation)软件包实现了混合法和随机抽样法两种不确定性分析方法,以研究多群核数据不确定性对堆芯物理计算的影响。以3×3假想堆芯为对象,对两种方法进行了验证,然后应用于国际原子能机构(International Atomic Energy Agency,IAEA)燃料管理基准题中的Almaraz核电厂首循环堆芯。分析结果表明,两种方法结果符合良好,Almaraz核电厂堆芯keff不确定性约为0.5%,堆芯径向和轴向功率的最大不确定性分别为1.9%和0.45%。     

HAZOP-study on heavy water research reactor primary cooling system

By knowledge-based Hazard and Operability (HAZOP) technique, equipment malfunction and deficiencies in the primary cooling system of the generic heavy water research reactor are studied. This technique is used to identify the representative accident scenarios. The related Process Flow Drawing (PFD) is prepared as our study database for this plant. Since this facility is in the design stage, applying the results of HAZOP-study to PFD improves the safety of the plant.     

Extended dryout and rewetting of small-particle core debris

Extended dryout of core debris, especially the location, size and temperature distribution of the dry zone and the variation of these parameters with time and power density (down to rewetting) was studied with volume-heated beds. The beds were composed of small, spherical stainless steel particles and water was used as coolant. In addition, transient rewetting was investigated starting with a dry bed of uniform temperature.     

Principles of reliability analysis methods applied to an emergency core cooling system

A reliability analysis is given for a model system of an emergency core cooling system of a pressurized water reactor. In order to demonstrate some basic relationships and influences on the system's failure probability the analysis deals only with some of the main-components and subsystems of the emergency core cooling system. With reference to the design basis accident, i.e. the total rupture of a main coolant line, only the low pressure system is considered. The overall system's failure probability is determined by the failure probability per demand, i.e. the unavailability of the system when called on for operation in the emergency case, and the cumulative probability of failure during the subsequent phase of residual heat removal. Detailed calculations have shown that the failure probability per demand is the leading term. Special attention is given to some parameter calculations dealing with the influence of inspection time intervals and repair procedures for different components and subsystems with respect to system failure behaviour.     

Thermal-fluid assessment of the design options for reactor vessel cooling in a prismatic core VHTR

The design of the reactor pressure vessel is an important issue in the VHTR design due to its high operating temperature. The extensive experience base in Light Water Reactor makes SA508/533 steel emerge as a strong candidate for the VHTR reactor vessel but requires maintaining the vessel temperature below the ASME code limit. To meet the temperature requirement, three types of vessel cooling options for a prismatic core VHTR are considered: an internal vessel cooling, an external vessel cooling, and an internal insulation. The performances of the vessel cooling options are evaluated by using a system thermo-fluid analysis code and a commercial computational fluid dynamics code during normal operation and accidents. The results suggested that the internal vessel cooling with the modified inlet flow path will be a promising option. The external cooling option does not ensure an effective cooling of the RPV. The insulation option provides an effective reduction of the RPV temperature in the normal and accident conditions but reduces the fuel safety margin during the accidents, requiring careful consideration before the implementation.     

Advanced methods in teaching reactor physics

Modern computer codes allow detailed neutron transport calculations. In combination with advanced 3D visualization software capable of treating large amounts of data in real time they form a powerful tool that can be used as a convenient modern educational tool for (nuclear power plant) operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for fuel management, core analysis and irradiation planning. The paper treats the visualization of neutron transport in different moderators, neutron flux and power distributions in two nuclear reactors (TRIGA type research reactor and typical PWR). The distributions are calculated with MCNP and CORD-2 computer codes and presented using Amira software.