Reactor Primary Circuit Research Articles

Detection of Neutrons Emitted From Reactor Primary Circuit Water by Discontinuing Flow Method

Abstract On-line activity measurement of fission products in a primary circuit water is often used for a fuel failure detection in research and power nuclear reactors. When gamma spectrometry is used for the activity measurement, high signal from 16N radionuclide and other activation products make the detection of fission products difficult. The detection of delayed neutrons emitted from several fission products is also used; however, if the detector is placed near the outlet coolant pipe, the signal from the delayed neutrons cannot be distinguished from the neutrons emitted due to 17N decay and deuterium photofission, with exception of a reactor scram condition. In this paper, a method of discontinuing the flow of primary circuit water is described. This method is based on the water flowing through a bypass on the outlet pipe to the sampling container and the flow is periodically temporarily interrupted, e.g., using 200 s + 200 s cycles. Neutrons located in the vicinity of the sampling container are continuously detected with a measuring sampling time of less than 2 s. The signal part, corresponding to the delayed neutrons, is evaluated by the signal decay analyzing during the flow interruption. The main sources of delayed neutrons suitable for this method are 137I, 87Br, and 88Br radionuclides with half-lives of 24.5 s, 55.7 s, and 16.5 s, respectively. The method was theoretically analyzed and experimentally verified in the LVR-15 research reactor.

Conceptual issues of the cold filter trap development for the sodium coolant purification in fast-neutron reactors

The paper presents the results of studying the peculiarities of heat and mass exchange in cold traps for the sodium purification of impurities in fast reactor circuits both in dedicated test areas simulating various trap components (isothermal sump, nonisothermal sump, filters, final cooling area) and in trap prototype models. As a result, a scientific rationale has been formed for developing traps of a unique design for various reactors. The impurity capacity of the traps is three to four times as high as that of the best foreign counterparts. Tests have shown these to be highly efficient in purifying sodium of oxygen and hydrogen and much less efficient in sodium purification of corrosion products and carbon. Taking into account the leakage of radioactive sodium during operation of the BN-600 reactor primary circuit traps, a decision was made to install the purification system in the reactor tank to improve the safety of the large fast reactor. It was resolved to exclude the accumulation of hydrogen in the primary circuit traps in nominal conditions. Two trap designs, with argon and sodium cooling, are discussed. It has been shown that operation of the reactor purification system with argon cooling will require 20 trap replacements during the reactor operating life and seven replacements if the deposition of hydrogen into the primary circuit cold traps is excluded. The sodium-cooled version of the trap built in the reactor tank has the same overall dimensions as the argon-cooled trap. The cooling sodium circulates in two trains: outside the jacketed working space body (up to 30% of the flow rate) and in the coil inside of the working space (up to 70% of the flow rate). Updates have been proposed to the trap design based on the calculations using the codes simulating the in-trap processes of heat and mass exchange.

Code development on steady-state thermal-hydraulic for small modular natural circulation lead-based fast reactor

Small Modular Reactors (SMRs) are attracting wide attention due to their outstanding performance, extensive studies have been carried out for lead-based fast reactors (LFRs) that cooled with Lead or Lead-bismuth (LBE), and small modular natural circulation LFR is one of the promising candidates for SMRs and LFRs development. One of the challenges for the design small modular natural circulation LFR is to master the natural circulation thermal-hydraulic performance in the reactor primary circuit, while the natural circulation characteristics is a coupled thermal-hydraulic problem of the core thermal power, the primary loop layout and the operating state of secondary cooling system etc. Thus, accurate predicting the natural circulation LFRs thermal-hydraulic features are highly required for conducting reactor operating condition evaluate and Thermal hydraulic design optimization. In this study, a thermal-hydraulic analysis code is developed for small modular natural circulation LFRs, which is based on several mathematical models for natural circulation originally. A small modular natural circulation LBE cooled fast reactor named URANUS developed by Korea is chosen to assess the code's capability. Comparisons are performed to demonstrate the accuracy of the code by the calculation results of MARS, and the key thermal-hydraulic parameters agree fairly well with the MARS ones. As a typical application case, steady-state analyses were conducted to have an assessment of thermal-hydraulic behavior under nominal condition, and several parameters affecting natural circulation were evaluated. What's more, two characteristics parameters that used to analyze natural circulation LFRs natural circulation capacity were established. The analyses show that the core thermal power, thermal center difference and flow resistance is the main factors affecting the reactor natural circulation. Improving the core thermal power, increasing the thermal center difference and decreasing the flow resistance can significantly increase the reactor mass flow rate. Characteristics parameters can be used to quickly evaluate the natural circulation capacity of natural circulation LFR under normal operating conditions.

A study into the modes of the VVER-1000 RCP starting in an earlier inoperative loop

To simulate the mode of the RCP starting in an earlier inoperative loop, KORSAR/GP, a code supporting coupled numerical modeling of neutronic and thermal-hydraulic transients in a VVER reactor plant in operating and emergency conditions, was chosen as the computational tool. Studying these modes using thermal-hydraulic codes makes it possible to analyze the course of transients and certain emergency processes without using commercial test procedures, which contributes to laying the groundwork for addressing the issues involved in ensuring the reliability, operating safety and efficiency of nuclear power plants. Increased requirements to the safety of NPPs identify the need for avoiding excessive conservatism in the analysis based on which requirements to safety systems are formulated, as well as for enhancing the knowledge of the regularities of thermal-hydraulic transients based on advanced computer programs (or codes) designed for improved computational analysis of non-stationary thermal hydraulics in the water-cooled reactor circulation circuits in emergency and transient modes relying on inhomogeneous non-equilibrium mathematical models of two-phase flows and on a detailed description of the physical transient regularities. The purpose of the study is to analyze computationally the starting of a VVER-1000 RCP in an earlier inoperative loop at different reactor plant power values. To do this, one requires to develop the VVER-1000 reactor primary circuit computational pattern to model the transient taking place as one RCP is started, to conduct a further analysis, and to compare the key monitored reactor coolant and core parameters (power, temperature, flow rate, etc.).

Study on bidirectional fluid-solid coupling characteristics of reactor coolant pump under steady-state condition

The AP1000 reactor coolant pump is a vertical shielded-mixed flow pump, is the most important coolant power supply and energy exchange equipment in nuclear reactor primary circuit system, whose steady-state and transient performance affect the safety of the whole nuclear island. Moreover, safety demonstration of reactor coolant pump is the most important step to judge whether it can be practiced, among which software simulation is the first step of theoretical verification. This paper mainly introduces the fluid-solid coupling simulation method applied to reactor coolant pump, studying the feasibility of simulation results based on workbench fluid-solid coupling technology. The study found that: for the unsteady calculations of the pure liquid media, the average head of the reactor coolant pump with bidirectional fluid-solid coupling decreases to a certain extent. And the coupling result is closer to the real experimental value. The large stress and deformation of rotor under different flow conditions are mainly distributed on impeller and idler, and the stress concentration mainly occurs at the junction of front cover plate and blade outlet. Among the factors that affect the dynamic stress change of rotor, the pressure load takes a dominant position.

Механизмы и главные причины повреждений тепловыделяющих элементов и тепловыделяющих сборок атомных электрических станций с реакторами с водой под давлением

This article is the first part of the review of publications on the problem of increasing the service life and improving the reliability of fuel assemblies (FA) and fuel rods (FR). In the next issues of the journal, Part 2 Dynamic interactions between the coolant flow and FAs and Part 3 Prediction of vibroacoustic resonances in the reactor cores of VVER-based nuclear power of this review will be published. This review part presents the results from investigations into the causes and mechanisms of FR failures, in which radioactive materials (gaseous fission products and volatile elements, in particular, krypton, xenon, iodine, and cesium) escape from the fuel pellet into the reactor primary coolant circuit. It is shown that the nuclear fuel depressurization level is influenced by all stages of the product life cycle, including design, manufacture, and operation of FAs. In particular, during operation, FA vibrations have an effect on the nuclear fuel depressurization level. As is well-known, an FA is a hydroelastic oscillatory system consisting of two interacting subsystems: a mechanical one and a hydrodynamic one. The FA vibration is due to the dynamic processes occurring in these subsystems. It is noted that the hydrodynamic subsystem has been studied to a rather insufficient extent, which is due to complexity of describing the processes through which the random hydrodynamic loads applied to the streamlined surfaces are produced, as well as the influence of flow thermal-hydraulic and acoustic characteristics on the occurrence of self-excited oscillations. Based on the defective FA operation data, the service life of leaky fuel rods and their linear heat generation rates are estimated. The evolutions of fuel operation conditions and fuel design changes are described. The available recommendations on optimizing FA designs developed taking into account the experience gained from operation of the reactors and the results of the accomplished fretting wear investigations are mentioned. It is shown that all who manufacture and operate nuclear fuel take efforts aimed at improving its quality, because it is the key indicator characterizing the vendor of fuel and nuclear power plants in the world market. Achieving longer fuel campaigns and higher burnup values while keeping or even improving the safety level are the most important criteria for nuclear fuel quality. Currently, the project called Nuclear Fuel Zero Failure Level is in the stage of implementation and obtaining the first results of joint activities of all participants.

Assessment of radioactive contamination in primary circuit of WWER-440 type reactors by computer code OSCAR for the decommissioning case

The article presents the results of modeling of the contamination of primary circuit of WWER-440 type reactor. The modeling has been performed using the OSCAR computer code (Nuclear research Centre in Cadarache, CEA, France), version V1.3 for decommissioning of nuclear power plants, taking into account the peculiarities of the WWER-440 reactor primary circuit, such as geometry, volumes, operational regime, materials. The results have been analyzed in the light of the scaling factor approach with the aim to demonstrate the modeling capabilities to identify the possibility of scaling factor application and to reproduce the activity correlation between difficult to measure and key nuclides. Calculated scaling factors for corrosion products are comparable with the ones of other nuclear power plants. Calculation results of surface activity contamination confirmed the applicability of 60Co as a key radionuclide.

Corrosion of cast Stellite-3 analogue in simulated PWR conditions

Wear resistant alloys are required for deployment in pressurised water reactor primary circuits and in this context, the corrosion behaviour of a cast cobalt-based alloy following exposure for 30 days in lithiated water at 300 °C was investigated. Corrosion of the surface was observed, resulting in the formation of a ∼100 nm thick chromium- and silicon-rich oxide. Preferential corrosion of the matrix at its interface with just one of the carbide types was observed to a depth of ∼1 μm; for the first time this was shown to be not due to any inhomogeneity in the matrix but instead to be an electrochemical effect.

Leak-before-break analyses of PWR and BWR piping concerning size effects

Components of primary circuit piping of light water reactor (LWR) plants fulfil high safety requirements. Beside prevention against any damage, cracking, and leaking, pipes are especially designed to exclude double-ended ruptures, which may cause a high loss of coolant. One option in this strategy, supported by regulations and standards in several countries, is the leak-before-break (LBB) behavior: a crack-like leakage should be detected before it becomes a break.Evidently, many factors influence the LBB behavior of a specific piping section. Material properties, plant operation parameters, assumed accident scenarios and the local geometry determine the result of a LBB analysis as well as the associated safety margins. Therefore, an analysis is traditionally done case by case, and site by site, according to the applicable standard.In this article, typical components of pressure water reactor (PWR) and boiling water reactor (BWR) primary circuits are examined, and their LBB behavior is investigated. The German standard (KTA) and the US approach (U.S.NRC) are compared. The quantitative results are obtained with the GRS codes PROST and WinLeck, which include different methods for fracture assessment and leak rate calculation. A clear trend concerning the pipe size is found, resulting in large safety margins for large-diameter piping.

Reynolds stress turbulence model applied to two-phase pressurized thermal shocks in nuclear power plant

Nuclear power plants are subjected to a variety of ageing mechanisms and, at the same time, exposed to potential pressurized thermal shock (PTS) – characterized by a rapid cooling of the internal Reactor Pressure Vessel (RPV) surface. In this context, NEPTUNE_CFD is used to model two-phase PTS and give an assessment on the structural integrity of the RPV. The first available choice was to use standard first order turbulence model (k-ɛ) to model high-Reynolds number flows encountered in Pressurized Water Reactor (PWR) primary circuits. In a first attempt, the use of k-ɛ model did produce some satisfactory results in terms of condensation rate and temperature field distribution on integral experiments, but also highlights some weaknesses in the way to model highly anisotropic turbulence. One way to improve the turbulence prediction – and consequently the temperature field distribution – is to opt for more advanced Reynolds Stress turbulence Model. After various verification and validation steps on separated effects cases – co-current air/steam-water stratified flows in rectangular channels, water jet impingements on water pool free surfaces – this Reynolds Stress turbulence Model (Rij-ɛ SSG) has been applied for the first time to thermal free surface flows under industrial conditions on COSI and TOPFLOW-PTS experiments. Coupled with the Large Interface Model, the first results confirmed the adequacy and increased accuracy of the approach in an industrial context.

Evaluation of operability of WWR-M reactor primary circuit piping with welded joint defects

Evaluation of operability of WWR-M reactor primary circuit piping with welded joint defects

On the shape of stress corrosion cracks in sensitized Type 304 SS in Boiling Water Reactor primary coolant piping at 288 °C

Evolution of the shape of surface cracks in sensitized Type 304 SS in Boiling Water Reactor primary coolant circuit piping at the reactor operating temperature of 288°C is explored as a function of various environmental variables, such as electrochemical potential (ECP), solution conductivity, flow velocity, and multiplier for the oxygen reduction reaction (ORR) standard exchange current density (SECD), using the coupled environment fracture model (CEFM). For this work, the CEFM was upgraded by incorporating Shoji’s model for calculating the crack tip strain rate and more advanced expressions were used for estimating the stress intensity factor for semi-elliptical surface cracks. This revised CEFM accurately predicts the dependence of the crack growth rate on stress intensity factor and offers an alternative explanation for the development of semi-elliptical cracks than that provided by fracture mechanics alone. The evolution of surface crack semi-elliptical shape depends strongly upon various environmental variables identified above, and the CEFM predicts that the minor axis of the ellipse should be oriented perpendicular to the surface, in agreement with observation. The development of the observed semi-elliptical cracks with the minor axis perpendicular to the surface is therefore attributed to the dependence of the crack growth rate on the electrochemical crack length.

SCC in PWRs: Learning from a Bottom-Up Approach

Stress corrosion cracking (SCC) of steels and Ni-base alloys in the pressurized water reactor (PWR) primary circuit has been a cause for reactor outages for many decades. Although the nuclear industry has made a considerable research effort to understand and predict SCC in this system, this enterprise is complicated by the sheer number of interdependent variables that have a major influence on the degradation behavior. SCC is highly time-dependant and often only sets in after many years. Therefore, autoclave testing, even with accelerating conditions, is an expensive and time-consuming endeavor. However, the results collected by a great number of research groups over many years have identified the most important parameters and, in many cases, how they influence the degradation behavior. The community has been constantly working on the development of a general theory that encompasses the underlying mechanisms and is capable of describing and predicting the observed degradation behavior. In the last two decades, the focus of research has shifted from the traditional approach of autoclave testing (for susceptibility and crack growth rates) to high-resolution microscopy and chemical analysis. The newly available techniques are providing data on chemical and structural changes locally at the crack tip where SCC occurs. The techniques covered in this review can provide very high chemical sensitivity at atomic resolution, which is ultimately needed in the quest for a generalized theory.

A phenomenological analysis of melt progression in the lower head of a pressurized water reactor

The analysis of in-vessel corium cooling (IVC) and retention (IVR) involves the description of very complex and transient physical phenomena. To get round this difficulty, “bounding” situations are often emphasized for the demonstration of corium coolability, by vessel flooding and/or by reactor pit flooding. This approach however comes up against its own limitations. More realistic melt progression scenarios are required to provide plausible corium configurations and vessel failure conditions.Work to develop more realistic melt progression scenarios has been done at CEA, in collaboration with EDF. Development has concentrated on the French 1300 MWe PWR, considering both dry scenarios and the possibility of flooding of the RPC (reactor primary circuit) and/or the reactor pit.The models used for this approach have been derived from the analysis of the TMI2 accident and take benefit from the lessons derived from several programs related to pool thermal hydraulics (BALI, COPO, ACOPO, etc.), material interactions (RASPLAV, MASCA), critical heat flux (CHF) on the external surface of the vessel (KAIST, SULTAN, ULPU), etc.Important conclusions of this work are as follows:(a)After the start of corium melting and onset of melt formation in the core at low pressure (∼1 to 5bars), it seems questionable that RPV (reactor pressure vessel) reflooding alone would be sufficient to achieve corium retention in the vessel;(b)If the vessel is not cooled externally, it may fail due to local heat-up before the whole core fuel inventory is relocated in the lower head;(c)Even if the vessel is externally cooled, transient 3D corium configurations in the lower head can induce high thermal loads on the vessel leading to a fast failure. Early in-vessel reflooding may prevent this risk. As a consequence, ex-vessel flooding should be accompanied by in-vessel reflooding;(d)In case of failure with external cooling, the breach is plausibly located on the lateral surface of the lower head. The failure is most likely local.The present work identifies effects that should be taken into account and modelled in numerical tools describing core degradation and vessel failure.

Experimental studies in water for safety grade decay heat removal of prototype fast breeder reactor

Decay heat removal is an important safety feature of any nuclear reactor. Prototype Fast Breeder Reactor (PFBR) is sodium cooled pool type reactor under construction at Kalpakkam, India. PFBR consists of two decay heat removal systems i.e. Operational Grade Decay Heat Removal (OGDHR) system and Safety Grade Decay Heat Removal (SGDHR) system. SGDHR system is a passive decay heat removal system based on natural circulation. Heat removal by natural convection is governed by many factors such as reactor configuration, and operating conditions. Hence it is essential to understand the thermal hydraulics of decay heat removal by natural circulation. In this regard experimental studies were carried out in 1/4th scale model of the reactor primary circuit using water as simulant to demonstrate the effectiveness of the Decay Heat Exchangers (DHX) on core cooling by establishment of the natural circulation in the model. This will help to understand decay heat removal by natural circulation and temperature pattern in the reactor during decay heat removal operation. Heat removal by inter wrapper flow is very effective during SGDHR operation. System is able to achieve steady state with only two DHX in operation at elevated temperature. Temperature pattern in the hot pool is lesser stratified with only two DHX operation. This paper discusses about similarity criteria followed, details of the experimental model, description of the instrumentation and experimental methodology and the experimental results.

A preliminary approach to the MSFR control issues

The Molten Salt Fast Reactor (MSFR) has been recognised as a promising reactor concept and has been adopted as reference circulating-fuel MSR configuration in the framework of the Gen-IV International Forum. Up to now, a number of dynamic studies have been performed to investigate the transient response of the system and its stability, but no control-oriented analyses have been carried out yet.This work focuses on a preliminary evaluation of the MSFR controllability issues, assuming a decentralized feedback scheme of control. For this purpose, a simplified model of the reactor primary circuit has been set up adopting a point-kinetics description and a one-dimensional description of neutronics and thermal–hydraulics, respectively. First, the effect of all the possible input variables on different sets of output variables is investigated by means of the Relative Gain Array (RGA) analysis, providing a brief overview of the best input/output pairings with no restrictions on the number of control variables that can be used. Some control schemes are then selected and implemented in the model of the primary circuit. The transient responses are analysed and compared, considering different control strategies.This work is thought to provide a first insight into some MSFR controllability issues as well as a preliminary evaluation of possible control options.

Design performances and chemistry program supporting the FA3/UK-EPRTMactivity management: experience and modelling balance

EPRTM reactor accounts with an evolutionary design that provides the appropriate features to ensure the safety implementation of different chemistry and radiochemistry options. ALARP considerations have been taken into account by EDF-AREVA for making decisions relating to the activity management in the primary circuit of FLAMANVILLE 3-EPRTM and UK-EPRTM reactors.The water chemistry and radiochemistry concept implemented in FA3-EPRTM and UK-EPRTM reactors is the result of an exhaustive selection process based on the balance between the theoretical developments, the laboratory tests and the NPP experience concerning the diverse areas associated with:• The source term identification and characterization: The understanding of the origin and behavior of fission products/actinides, corrosion products and activation products constitutes the essential support for the selection of suitable parameters and criteria to monitor the system integrity, the tramp-uranium and radiation build-up and the discharges to the environment.• The source term quantification: The balance between the baseline data from PWR forerunner reactors and the assessments performed by modelling constitutes the major demonstration of the source term accuracy. This approach ensures that activity risks are understood and can be managed with the EPRTM design options.The EPRTM design options evaluation: The sensitivity analysis results show the influence of the fuel management, the material choice and the chemistry conditioning on several domains such as the activity coolant and the fuel/ex-core crud management.EDF-AREVA demonstrates by means of this process that the design, sizing and chemistry conditioning of EPRTM reactor primary circuit are adapted to guarantee the correct activity management. The methodology developed, based on qualitative and quantitative assessments, intends to propose to the Nuclear Industry several alternatives for evaluating and/or improving the compliance with requirements associated with safety, material integrity, radioprotection and environment.

Natural convection test in Phenix reactor and associated CATHARE calculation

The Phenix sodium cooled fast reactor (SFR) started operation in 1973 and was stopped in 2009. Before the reactor was definitively stopped, ultimate tests were performed, including a natural convection test in the primary circuit. One objective of this natural convection test is the validation of plant dynamic codes as CATHARE code for future safety studies on SFRs.The paper firstly describes the Phenix pool type reactor primary circuit with the main components and the instrumentation used during the tests. The initial test conditions and the detailed transient scenario are presented: steam generators dry out, scram, stop of the primary pumps, development of natural convection in the primary circuit with two different phases. Then, CATHARE modelling of the Phenix primary circuit is described. The whole test is calculated for a total duration of 7h in natural convection regime. The CATHARE calculations are compared to the Phenix measurements. A particular attention is paid to the significant decrease of the core power before the scram, due to the increase of temperature at the core inlet. Then, the computed evolution of main components inlet and outlet temperatures is compared to the reactor data. The need of coupling system code with CFD code to model the 3D behaviour of large pools during natural convection regime is pointed out.

Assessment of occupational internal exposure to beta emitters from the nuclear reactor primary coolant circuit

Fission products of 235U or isotopes from activation may appear in technological waters at normal operation of a research reactor. Therefore, reactor cooling water may contain a number of beta radioactive isotopes including yttrium and strontium isotopes, which can pose potential hazard to reactor personnel. In order to asses internal exposure urinalysis is carried out. This work presents the method of sample preparation and measurement used by Radiation Protection Measurements Laboratory (RPLM) of the National Centre for Nuclear Research (NCNR). Method of various isotopes of yttrium and Sr-90 activity calculation is also shown. Determination of these isotopes activities in urine sample allows estimating the effective doses

Behavior of stainless steels in pressurized water reactor primary circuits

Stainless steels are widely used in primary circuits of pressurized water reactors (PWRs). Operating experience with the various grades of stainless steels over several decades of years has generally been excellent. Nevertheless, stress corrosion failures have been reported in few cases. Two main factors contributing to SCC susceptibility enhancement are investigated in this study: cold work and irradiation.Irradiation is involved in the stress corrosion cracking and corrosion of in-core reactor components in PWR environment. Irradiated assisted stress corrosion cracking (IASCC) is a complex and multi-physics phenomenon for which a predictive modeling able to describe initiation and/or propagation is not yet achieved. Experimentally, development of initiation smart tests and of in situ instrumentation, also in nuclear reactors, is an important axis in order to gain a better understanding of IASCC kinetics.A strong susceptibility for SCC of heavily cold worked austenitic stainless steels is evidenced in hydrogenated primary water typical of PWRs. It is shown that for a given cold-working procedure, SCC susceptibility of austenitic stainless steels materials increases with increasing cold-work. Results have shown also strong influences of the cold work on the oxide layer composition and of the maximum stress on the time to fracture.