Primary Coolant Loop Research Articles

Simulation of fuel dispersion in the MYRRHA-FASTEF primary coolant with CFD and SIMMER-IV

The objective of this work is to assess the behavior of fuel redistribution in heavy liquid metal nuclear systems under fuel pin failure conditions. Two different modeling approaches are considered using Computational Fluid Dynamics (CFD) codes and a system code, applied to the MYRRHA facility primary coolant loop version 1.4.Two different CFD models are constructed: the first is a single-phase steady model prepared in ANSYS Fluent, while the second is a two-phase model based on the volume of fluid (VOF) method in STARCCM+ to capture the upper free-surface dynamics. Both use a Lagrangian tracking approach with oneway coupling to follow the particles throughout the reactor. The system code SIMMER-IV is used for the third model, without neutronic coupling. Although limited regarding the fluid dynamic aspects compared to the CFD codes, comparisons of particle distributions highlight strong similarities despite quantitative discrepancies in the size of fuel accumulations. These disparities should be taken into account while performing the safety analysis of nuclear systems and developing strategies for accident mitigation.

Reduced-scale water test of natural circulation for decay heat removal in loop-type sodium-cooled fast reactor

Water tests of a loop-type sodium-cooled fast reactor have been conducted to physically evaluate the natural circulation characteristics. The water test apparatus was manufactured as a 1/10-scale mock-up of the Japan Sodium-Cooled Fast Reactor, which adopts a decay heat removal system (DHRS) utilizing natural circulation. Tests simulating a variety of events and operation conditions clarified the thermal hydraulic characteristics and core-cooling performance of the natural circulation in the primary loop. Operation conditions such as the duration of the pump flow coast-down and the activation time of the DHRS affect the natural circulation characteristics. A long pump flow coast-down cools the upper plenum of the reactor vessel (RV). This causes the loss of the buoyant force in the RV. The test result indicates that a long pump flow coast-down tends to result in a rapid increase in the core temperature because of the loss of the buoyant force. The delayed activation of the DHRS causes a decrease in the natural circulation flow rate and a temperature rise in the RV. Flow rate deviation and a reverse flow appear in the parallel cold-leg piping in some events, which cause thermal stratification in the cold-leg piping. The DHRS prevents the core temperature from fatally rise even for the most severe design-basis event, in which sodium leakage in a secondary loop of the DHRS and the opening failure of a single damper of the air cooler occur simultaneously. In the water test for the case of siphon break in the primary loop, which is one of the design extension conditions, a circulation flow consisting of ascendant and descendant flows in the core channels suppresses the rise in the core temperature instead of the primary coolant loop. However, the test result suggests that the present DHRS has insufficient performance to remove the decay heat in the case of siphon break. Repeated tests have shown that the results of the water test have sufficiently high reproducibility in terms of the flow rate of the primary loop to be used for the verification of numerical codes.

Parametric Study of Time-Dependent Corrosion Product Activity due to56Mn,58Co, and60Co in the Primary Coolant Circuit of a Typical Pressurized Water Reactor

Results of a detailed study, based on the parametric analysis of activated corrosion products, in primary coolant of a typical pressurized water reactor (PWR) are presented. The parameters influencing time dependent buildup of corrosion product activity (CPA) in primary coolant loop of PWR were identified. The computer program CPAIR was used to accommodate for time dependent corrosion rates. The behaviors of56Mn,58Co, and60Co were studied over the reactor operational time. During the course of normal operation of reactor, the CPA is dominated by56Mn, while58Co and60Co are the predominant radionuclides after reactor shutdown. Parametric study suggests that the total CPA is most sensitive to ion-exchanger removal rates. For a removal rate of 300 cm3-s−1, the specific activity due to56Mn has the maximum value of 3.552 × 104 Bq-m−3after 1,000 hours of reactor operation. This value decreases drastically to 8.325 × 103 Bq-m−3at removal rate of 900 cm3-s−1. Additionally, CPA due to56Mn,58Co, and60Co shows strong dependence on removal rates from the core material surfaces. Variations in the values of radionuclide removal rates from piping surface and radionuclide removal rate from deposition on pipes showed only very small effects on CPA buildup.

DEVELOPMENT AND VALIDATION OF THE AEROSOL TRANSPORT MODULE GAMMA-FP FOR EVALUATING RADIOACTIVE FISSION PRODUCT SOURCE TERMS IN A VHTR

Predicting radioactive fission product (FP) behaviors in the reactor coolant system and the containment of a nuclear power plant (NPP) is one of the major concerns in the field of reactor safety, since the amount of radioactive FP released into the environment during the postulated accident sequences is one of the major regulatory issues. Radioactive FPs circulating in the primary coolant loop and released into the containment are basically in the form of gas or aerosol. In this study, a multi-component and multi-sectional analysis module for aerosol fission products has been developed based on the MAEROS model [1,2], and the aerosol transport model has been developed and verified against an analytic solution. The deposition of aerosol FPs to the surrounding structural surfaces is modeled with recent research achievements. The developed aerosol analysis model has been successfully validated against the STORM SR-11 experimental data [3], which is International Standard Problem No. 40. Future studies include the development of the resuspension, growth, and chemical reaction models of aerosol fission products.

Simulation of advanced accumulator and its application in CPR1000 LBLOCA analysis

The advanced accumulator is designed to improve the safety and reliability of CPR1000 by providing a small injection flow to keep the reactor core in flooded condition. Thus, the core still stays in a cooling state without the intervention of low pressure safety injection and the startup grace time of the low pressure safety injection pump can be greatly extended.A new model for the advanced accumulator, which is based on the basic conservation equations, is developed and incorporated into RELAP5/MOD 3.3. The simulation of the advanced accumulator can be carried out and results show that the behavior of the advanced accumulator satisfied its primary design target. There is a large flow in the advanced accumulator at the initial stage. When the accumulator water level is lower than the stand pipe, a vortex appears in the damper, which results in a large pressure drop and a small flow. And then the sensitivity analysis is performed and the major factors which affected the flow rate of the advanced accumulator were obtained, including the damper diameter, the initial volume ratio of the water and the nitrogen and the diameter ratio of the standpipe and the small pipe.Additionally, the primary coolant loop cold leg double-ended guillotine break LBLOCA in CPR1000 with advanced accumulator is analyzed. The results show that the criterion for maximum cladding temperature limit (1477K) (NRC, 1992) can be met ever with 200s after the startup of the low pressure safety injection. From this point of view, passive advanced accumulator can strive a longer grace time for LPSI. Thus the reliability, safety and economy of the reactor system can be improved.

Structural Safety Analysis Based on Seismic Service Conditions for Butterfly Valves in a Nuclear Power Plant

The structural integrity of valves that are used to control cooling waters in the primary coolant loop that prevents boiling within the reactor in a nuclear power plant must be capable of withstanding earthquakes or other dangerous situations. In this study, numerical analyses using a finite element method, that is, static and dynamic analyses according to the rigid or flexible characteristics of the dynamic properties of a 200A butterfly valve, were performed according to the KEPIC MFA. An experimental vibration test was also carried out in order to verify the results from the modal analysis, in which a validated finite element model was obtained via a model-updating method that considers changes in the in situ experimental data. By using a validated finite element model, the equivalent static load under SSE conditions stipulated by the KEPIC MFA gave a stress of 135 MPa that occurred at the connections of the stem and body. A larger stress of 183 MPa was induced when we used a CQC method with a design response spectrum that uses 2% damping ratio. These values were lower than the allowable strength of the materials used for manufacturing the butterfly valve, and, therefore, its structural safety met the KEPIC MFA requirements.

Elastic–plastic analysis of the steel-to-tungsten transition joint for a high performance divertor

The use of tungsten as a plasma-facing material necessitates a transition joint to the oxide dispersion strengthened (ODS) steel or ferritic steel (FS) structural material of the primary coolant loop at the end of the divertor target plate where the surface heat flux is very low. A critical issue in the transition joints is the coefficient of thermal expansion (CTE) mismatch between the tungsten (or tungsten-alloy) and ODS steel, which can lead to unacceptably high thermal stresses during steady state and ratcheting during cyclic loads. Detailed 2D and 3D thermo-mechanical analyses were conducted to study the behavior of a transition from tungsten to FS with an intermediate layer of tantalum, located outside of the high heat flux region. The results include plastic strains under various loading conditions including fabrication processes, warm and cold shutdown, and allow for plastic behaviors leading to stress relaxation. The accumulation of plastic deformation may cause ratcheting. Modifications were proposed to the transition joint design in order to eliminate stress concentration and ratcheting under cyclic loading. The results of the modified design exhibited less plastic deformation in the joints as well as no ratcheting caused by warm and cold shutdown.

Development of a transient thermal-hydraulic code for analysis of China Demonstration Fast Reactor

The transient thermal-hydraulic code THACOS is under development for analysis of China Demonstration Fast Reactor. Applying modular technology, the code contains the core module, the pump module, the sodium pool module and the heat exchanger module and each module could operate separately. It can provide one-dimensional thermal-hydraulic simulation for the primary sodium coolant loop. The point reactor kinetics equations with six-group delayed neutrons have been applied to calculate the core power considering reactivity feedbacks caused by the Doppler effect, coolant density, axial expansion of fuel rods and radial expansion of the core. Multiple-channel model is applied to depict the core. Compressible homogenous flow model is used for the two-phase flow of sodium. The calculated results show that sodium boiling will occur quickly under the ULOF accident without any shutdown rods insertion. While, with the insertion of three hydraulically suspended shutdown rods, the core could be shut down safely and boiling will not occur in a short period of time. Obviously, the passive hydraulically suspended shutdown rods could keep the core safe under ULOF accidents.

FASTEF Heat exchanger tube rupture CFD simulation

The aim of this technical note is to present CFD simulations of a tube rupture incidental scenario in a Primary Heat eXchanger (PHX)/Primary Pump (PP) assembly for two design variants of the FAst-Spectrum Transmutation Experimental Facility FASTEF ongoing design, in the framework of the FP7 Central Design Team (CDT) European project. The simulation domain reproduces with some simplification the entire primary coolant loop. The objective is to understand whether it is necessary take some counter-measures to avoid the ingress of steam in the cold plenum. The simulation has been performed on two successive updates of the design and of the nominal operation. The simulations show a good resistance to steam ingress, under the condition that provision is made to avoid an excessive accumulation of steam at the top of the PHX/PP assembly casing.

Research on Fluid-Structure Interaction Dynamic Characteristics of Steam Generator Heat Exchanger Tubes

For heat exchanger tube of steam generator, the relation between heat exchanger tube and fluid is typical fluid-structure interaction problem. Flow induced vibration has been found so far to be responsible for fatigue damage and failure of steam generator tubes, which will result in large economic loss and radioactive pollution. So the steam generator tubes are the weakest link in the primary coolant loop. Based on the synthesis of all sorts of factors influencing the dynamic characteristics of steam generator heat transfer tubes, establishing the heat transfer tube model, research on the weakening effect of fluid hole on fluid, the natural frequencies of the heat transfer tubes are analyzed under different fluid holes and fluid hole distance by numerical simulation.

Elastic-Plastic Analysis of the Transition Joint for a High Performance Divertor Target Plate

The use of tungsten as a plasma-facing material necessitates a transition joint to the structural material of the primary coolant loop at some location in order to transport the coolant to the heat exchanger. A critical issue in transition joints is the thermal expansion mismatch between materials, which can lead to unacceptably high thermal stresses. Detailed 2D and 3D analyses were performed to study the behavior of a transition from tungsten to ferritic steel (FS) with an intermediate layer of tantalum, located outside of the high heat flux region. This paper describes the results of FEM analyses including primary and secondary stresses under various time-dependent loading conditions such as warm and cold shutdown, and allowing for inelastic behaviors leading to stress relaxation and ratcheting. The results show that the transition joint satisfies the design requirement on maximum accumulated principal strain during operation.

Transient behaviour of a parallel pump in nuclear research reactors

A theoretical analysis of the fast transients of a parallel pump, based on inertia of the rotating parts and inertia of the fluid, is proposed. It leads to total torque, total head, and total system resistance during transient periods. The equations indicate that an increase in coolant inertia increases the acceleration head. While an increase in the moment of inertia of rotating parts decreases the acceleration head. The model is used to analyze the behaviour of the Tehran Research Reactor (TRR) primary coolant loop parallel pump during a fast startup. The results of present model are compared with similar studies and good agreement is observed.

NaK release model for MASTER-2009

Sodium–potassium droplets from the primary coolant loop of Russian orbital reactors have been released into space. These droplets are called NaK droplets. Sixteen nuclear powered satellites of the type RORSAT launched between 1980 and 1988 activated a reactor core ejection system, mostly between 900 and 950km altitude. The core ejection causes an opening of the primary coolant loop. The liquid coolant consists of eutectic sodium–potassium alloy and has been released into space during these core ejections. The NaK coolant has been forming droplets up to a diameter of 5.5cm. NaK droplets have been modeled before in ESA's MASTER Debris and Meteoroid Environment Model. The approach is currently revised for the MASTER-2009 upgrade. A mathematical improvement is introduced by substituting the current size distribution function by the modified Rosin–Rammler equation. A bimodal size distribution is derived which is based on the modified mass based Rosin–Rammler equation. The equation is modified by truncating the size range and normalizing over the finite range between the size limits of the smallest and the biggest droplet. The parameters of the model are introduced and discussed. For the validation of the NaK release model, sixteen release events are simulated. The resulting size distribution is compared with radar measurement data. The size distribution model fits well with revised published measurement data of radar observations. Results of orbit propagation simulation runs are presented in terms of spatial density.

A semi-empirical approach for predicting two-phase flow discharge through branches of various orientations connected to a horizontal main pipe

The subdivision of two-phase flow in branching conduits consisting of a large horizontal main pipe with upward, downward, or lateral branches of reduced diameter is of great interest in various technological fields. For example, these conduits are important in light-water nuclear reactors (LWRs) in the case of a small break loss-of-coolant accident (SBLOCA) in a leg of the reactor's primary coolant loops, as well as for breaks or valve malfunctions in a large pipeline. In these kinds of circumstances, the relevant phenomenology often involves phase stratification coupled with possible liquid entrainment or gas pool-through phenomena. Therefore, these phenomena were studied in depth to evaluate the pressure drop across conduit elements resulting from the redistribution of flow phases and the discharged mass among them. In the past, several experiments have been performed along with studies in flow modelling. As a result, several formulae and models for branch exit quality and consequent discharge mass flow rate predictions have been proposed. In recent years, we have been engaged in extensive research on this subject, resulting in a new semi-empirical formulation to express branch exit quality in terms of the physical processes taking place in the conduits, the operating parameters and the branch geometry. In this paper, we applied these correlations to representative sets of experiments previously reported in the literature, comprising a wide range of branch-to-main-pipe diameter ratios and operating pressures, which proved our formulation to be very accurate.

Estimation of the Doppler Coefficient from a Lower Power Transient Observed in a Zero-Power Reactor Physics Test of a PWR (I)—Methodology—

A method for estimating the Doppler coefficient from low power transient data that are measured in a routine experiment conducted in a zero-power reactor physics test of a PWR power plant has been developed. A dynamics identification model was introduced by considering two dynamics components for reactor kinetics and heat removal in the primary coolant loops. The dynamics identification method in the time domain was applied under consideration that the effect of reactivity feedback observed in the experiment was very small and measured transient data contained some intermittent intervals unavailable for dynamics identification. Unknown parameters, including the Doppler coefficient, contained in the dynamics model were determined stepwise through five estimation steps. In these steps, gamma ray noises were removed from the NIS signal, the reactivity feedback component was extracted from reactivity transient, and fuel temperature transient was determined in such way as to reproduce the observed low power transient in the digital simulation. In the final step, the Doppler coefficient was estimated from the extracted reactivity feedback component using the quantities evaluated in the preceding steps. The estimated value agreed well with the value evaluated with a core design code.

Estimation of the Doppler Coefficient from a Lower Power Transient Observed in a Zero-Power Reactor Physics Test of a PWR (I) —Methodology—

A method for estimating the Doppler coefficient from low power transient data that are measured in a routine experiment conducted in a zero-power reactor physics test of a PWR power plant has been developed. A dynamics identification model was introduced by considering two dynamics components for reactor kinetics and heat removal in the primary coolant loops. The dynamics identification method in the time domain was applied under consideration that the effect of reactivity feedback observed in the experiment was very small and measured transient data contained some intermittent intervals unavailable for dynamics identification. Unknown parameters, including the Doppler coefficient, contained in the dynamics model were determined stepwise through five estimation steps. In these steps, gamma ray noises were removed from the NIS signal, the reactivity feedback component was extracted from reactivity transient, and fuel temperature transient was determined in such way as to reproduce the observed low power transient in the digital simulation. In the final step, the Doppler coefficient was estimated from the extracted reactivity feedback component using the quantities evaluated in the preceding steps. The estimated value agreed well with the value evaluated with a core design code.

Leak Before Break application for Primary Coolant Loop and Surge Line of VVER-1000/320 Plant: TACIS Project R2.09/96

Activities were carried out within the frame of TACIS-96 Program—Project R2.09/96 named “LBB Applicability Review and Basic Implementation Engineering for Primary Coolant Loop & Surge Line of VVER-1000/320 Nuclear Power Plant”. Main objective of the Project was to perform a Leak Before Break assessment for the Main Coolant Pipes and Pressurizer Surge Line of the Reactor Coolant System of a selected “reference” Unit of a VVER-1000, type V-320 Nuclear Power Plant, which was assumed the Balakovo Unit 2. This has included the following: • definition of a “reference” methodology for primary piping systems with cladding; • verification of the Leak Before Break concept applicability to primary piping, in accordance with the “reference” methodology defined; • preparation of necessary materials data and validation files; • evaluation of effectiveness and reliability of the Leak Detection systems and the In-service Inspection Programs (methodologies and qualification); • preparation of a recommendations list to improve the existing information and diagnostic systems; • transfer of know–how from Consortium to Sub-Contractor, integrating the feedback of Western experience. One of the main highlights of the Project was the definition and execution of a comprehensive Material Testing Program in order to establish the actual tensile and fracture properties of the primary piping materials: base materials, weld metals, and “heat affected zones” were tested. Compact tests and standard specimens for J– R curves determination were used; special attention was paid to the dissimilar typical joints of the Surge Line pipes, being made by carbon steel base metal with stainless steel weld. Fracture properties as J– R curves were determined both at room and operating temperatures. The “reference” methodology defined has taken into account both Western and Russian Leak Before Break approaches. The analysis has shown that criteria for Leak Before Break concept applicability to the primary piping of Balakovo VVER-1000/320 Unit 2, presented in the reference methods developed in the frame of this Project, are met.

Boric acid recovery using polymer filtration: Studies with alkyl monool, diol, and triol containing polyethylenimines

Three water-soluble polymers containing linear alkyl monool, 1,2-diol, and 1,2,3-triol groups, mostly on the primary amines of polyethylenimine, were synthesized, characterized, and tested for their ability to recover boric acid. The boron-binding capacities of these polymers and the backbone polyethylenimine were determined by titration, ultrafiltration, and inductively coupled plasma/atomic emission spectroscopy analysis. At low boron concentrations, the 1,2,3-triol polymer performed better than the 1,2-diol, whereas at high boron concentrations, the 1,2-diol outperformed the 1,2,3-triol. 11B-NMR spectroscopy and retention studies with various salt concentrations indicated that boron interacted with these two polymers by means of ion pairing with the protonated amines and by borate ester formation. For the monool and the polyethylenimine backbone, the mechanism for boron binding was ion pairing only. These polymers are under consideration for the selective recovery and recycling of enriched boric acid used in the primary coolant loop of pressurized water nuclear reactors. © 2005 Wiley Periodicals, Inc. J Appl Polym Sci 97: 1590–1604, 2005

Thermal-mixing analyses for safety injection at partial loop stagnation of a nuclear power plant

When a cold HPSI (High Pressure Safety Injection) fluid associated with an overcooling transient, such as SGTR (Steam Generator Tube Rupture), MSLB (Main Steam Line Break) etc., enters the cold legs of a stagnated primary coolant loop, thermal stratification phenomena will arise due to incomplete mixing. If the stratified flow enters the downcomer of the reactor pressure vessel, severe thermal stresses are created in a radiation embrittled vessel wall by local overcooling. As general thermal-hydraulic system analysis codes cannot properly predict the thermal stratification phenomena, RG 1.154 requires that a detailed thermal-mixing analysis of PTS (Pressurized Thermal Shock) evaluation be performed. Also, previous PTS studies have assumed that the thermal stratification phenomena generated in the stagnated loop side of a partially stagnated primary coolant loop are neutralized in the vessel downcomer by the strong flow from the unstagnated loop. On the basis of these reasons, this paper focuses on the development of a 3-dimensional thermal-mixing analysis model using PHOENICS code which can be applied to both partial and total loop stagnated cases. In addition, this paper verifies the fact that, for partial loop stagnated cases, the cold plume generated in the vessel downcomer due to the thermal stratification phenomena of the stagnated loop is almost neutralized by the strong flow of the unstagnated loop but is not fully eliminated.

Analysis of an Advanced Test Reactor Small-Break Loss-of-Coolant Accident with an Engineered Safety Feature to Automatically Trip the Primary Coolant Pumps

A new engineered safety feature that automatically trips the primary coolant pumps following a low-pressure reactor scram was recently installed in the Advanced Test Reactor (ATR). The purpose of this engineered safety feature is to prevent the ATR’s surge tank, which contains compressed air, from emptying during a small-break loss-of-coolant accident (SBLOCA). If the surge tank were to empty, the air introduced into the primary coolant loop could potentially cause the performance of the primary and/or emergency coolant pumps to degrade, thereby reducing core thermal margins. Safety analysis performed with the RELAP5 thermal-hydraulic code and the SINDA thermal analyzer shows that adequate thermal margins are maintained during an SBLOCA with the new engineered safety feature installed. The analysis also shows that the surge tank will not empty during an SBLOCA even if one of the primary coolant pumps fails to trip.