Primary Heat Transport System Research Articles

EXPERIMENTAL RESEARCH CONCERNING WATERSIDE CORROSION OF FUEL CLADDING IN CANDU REACTOR

EXPERIMENTAL RESEARCH CONCERNING WATERSIDE CORROSION OF FUEL CLADDING IN CANDU REACTOR

Optimized Water Distillation Layout for Detritiation Purpose

Tritium permeation constitutes a key issue for the future EU-DEMO, especially in the Breeding Blanket (BB) where fusion energy must be delivered to the Primary Heat Transport System (PHTS) and where tritium must be bred. Currently, the mitigation strategy of the tritium permeation from BB into primary coolant is based on the adoption of anti-permeation barriers and on the operation of the Coolant Purification System (CPS). This system must ensure a tritium removal rate from the primary coolant equal to the BB permeation rate at a target tritium-specific activity inside the PHTS. In the case of the Water-Cooled Lithium Lead (WCLL) BB, water distillation was selected as the most promising technology for the primary coolant detritiation due to its intrinsic simplicity and safety. Nevertheless, power consumption was recognized as a relevant concern. For this reason, the present work aims at investigating possibilities to reduce power consumption of the water CPS implementing Heat Pump-Assisted Distillation (HPAD) concepts. To do this, a review of the HPADs developed in the chemical industry was carried out, and the best options for the water CPS were identified based on qualitative considerations. Then, a quantitatively assessment of the best solution in terms of power consumption and tritium inventory was performed with the commercial numerical tool Aspen Plus. Finally, the Mechanical Vapor Recompression (MVR) concept was recognized as the most promising solution, ensuring a power saving of around 80% while keeping a limited tritium inventory.

Activated corrosion product contamination assessments of DEMO WCLL breeding blanket primary heat transport system

In water-cooled fusion reactors, the assessment of the primary system contamination is essential for waste management, machine availability, occupational radiation exposure, and radiological hazard determination. The primary cooling water is not only directly activated by the intense neutron field but is a contamination vector for a significant variety of gamma emitters with short to long decay half-lives. Corrosion products can be activated in those regions under neutron flux of the primary circuit and then released in the cooling water.In the EU-DEMO fusion power plant equipped with the Water-Cooled Lithium Lead Breeding Blanket (WCLL-BB) concept, the primary coolant undergoes intense neutron fields in the first wall and the breeding zone regions of the blanket. Activated Corrosion Products (ACPs) are then formed, released into the water, transported in the cooling loop and finally deposited onto the ex -vessel surfaces of the Primary Heat Transport System (PHTS), where working personnel are susceptible to being radiologically exposed.This work addresses the complete assessment of ACPs in the WCLL-BB PHTS of EU-DEMO. The simultaneous and multi-physical processes behind the ACP formation are tackled using the OSCAR-Fusion code, a comprehensive tool developed by the CEA (France) to assess contamination in fusion nuclear reactors. The whole system is modelled with zero-dimensional nodes with assigned geometrical, thermal-hydraulics, material and chemical parameters. Activation reaction rates integrated over the whole spectrum and calculated with MCNP are given to those regions exposed to the neutron flux. Results are provided in terms of mass and activity inventories of ACPs as deposit and inner oxide layers of components (pipes, heat exchangers, pumps...), ions in solution, particles in suspension, and filters and resins trapping.Mobilizable inventories such as ions, particles and deposits are important source terms in accidental scenario evolutions, while the whole activity inventory constitutes the main long-term gamma emitting source for dose rate maps determination in the tokamak building rooms housing the main PHTS equipment.

Thermal-hydraulic study of the EU-DEMO Helium Cooled Pebble Bed Breeding Blanket Primary Heat Transport System

The European DEMO reactor will work under normal operating conditions in accordance with a pulsed duty cycle. However unplanned and planned transients of plasma overpower may occur compromising the integrity of its plasma-facing components structures. Consequently, adopting appropriate tools is essential to accurately and consistently model the thermal-hydraulic response of the involved cooling systems under both normal operating conditions and accidental events. Given this background, the University of Palermo, in collaboration with EUROfusion, started a research work to study the thermal-hydraulic behaviour of the Primary Heat Transport System (PHTS) of the Helium Cooled Pebble Bed Breeding Blanket (HCPB BB) of the DEMO device under steady-state and transient conditions. The activity has been performed adopting a computational approach, employing the thermal-hydraulic system code TRACE version 5.0 patch 6. The key point of the work has been the code-to-code benchmark with the outcomes previously obtained with the RELAP5-3D code, to estimate the impact of the physical models, numerical resolution schemes and modelling techniques adopted on the predictive capabilities of the system codes considered. The models and the analysis results are presented and critically discussed herein.

Fracture Behaviour of Carbon Steel Straight Pipe with Through-Wall Crack under Cyclic Loading

Carbon steel straight pipes form part of the primary heat transport system of nuclear power plants. They may be subjected to dynamic-cyclic loads during earthquake event. This dynamic cyclic loading significantly affects the fracture resistance and the load carrying capacity of cracked components. Further, for the application of leak-before-break concepts, experimental investigations are essential. Therefore, the aim of the present study is to investigate the effect of cyclic loading on fracture resistance of through-wall cracked straight pipes. The paper presents details of the cyclic fracture test on a straight pipe and the results thereof. The behaviour of the pipe was compared with the fracture behaviour under monotonic loading. The study has shown that the load capacity of the pipe is reduced under cyclic loading and the failure is by cyclic tearing followed by sudden collapse.

Iron hydrolysis and lithium uptake on mixed-bed ion exchange resin at alkaline pH

The use of ion exchange resins to remove ionic impurities from solution is prevalent in industrial process systems, including in the primary heat transport system (PHTS) purification circuit of nuclear power plants. Despite its extensive use in the nuclear industry, our general understanding of ion exchange cannot fully explain the complex chemistry in ion exchange beds, particularly when operated at or near their saturation limit. This work investigates the behaviour of mixed-bed ion exchange resin, saturated with species representative of corrosion products in a CANDU (Canadian Deuterium Uranium) reactor PHTS, particularly with respect to iron chemistry in the resin bed and the removal of lithium ions from solution. Experiments were performed under deaerated conditions, analogous to normal PHTS operation. The results show interesting iron chemistry, suggesting the hydrolysis of cation resin bound ferrous species and the subsequent formation of either a solid hydrolysis product or the soluble, anionic Fe(OH)3-.

Thermal-hydraulic study of the Primary Heat Transport System of the DEMO Divertor Cassette Body

The EU-DEMO reactor is supposed to undergo a pulsed duty cycle under normal conditions, which might challenge the qualified lifetime of the main equipment inducing undue thermal and mechanical cycling loads. Moreover, during power operation it is not possible to exclude the occurrence plasma overpower transients that might jeopardize the structural integrity of the plasma-facing components. In this context, the University of Palermo in cooperation with ENEA has launched a research campaign within the framework of the EUROfusion action to investigate the thermal-hydraulic behaviour of the Primary Heat Transport System (PHTS) of the EU-DEMO Divertor (DIV) Cassette Body (CB) during the typical DEMO duty cycle. The study was primarily intended to start the definition of the strategies to be followed to minimize thermal fluctuations due to the pulsed operation. In particular, the attention has been focussed on the assessment of the DIV CB PHTS temperature behaviour during the pulse-dwell transition so to check the potential effectiveness of a direct coolant temperature control by properly tuning the feedwater mass flow rate. The activity has been led following a computational approach based on the adoption of the TRACE thermal-hydraulic system code. Models, assumptions, and outcomes of the analyses are herewith reported and critically discussed.

Development of a thermal-hydraulic model of the EU-DEMO Water Cooled Lithium Lead Breeding Blanket Primary Heat Transport System

The EUROfusion consortium is developing the project of a DEMOnstration Fusion Reactor (EU-DEMO) which would follow ITER in the pathway towards the quest for the exploitation of fusion energy. EU-DEMO has been conceived to deliver net electric power to the grid. Therefore, proper critical evaluations of the tokamak cooling and power conversion systems are needed because they play a pivotal role in the design and licencing of the overall plant. The EU-DEMO reactor will be based on the tokamak concept and, as such, it is supposed to undergo a pulsed duty cycle under normal conditions, which might challenge the qualified lifetime of the main equipment inducing undue thermal and mechanical cycling. Moreover, the EU-DEMO plasma control strategy postulates the possible occurrence of planned and off-normal plasma overpower transients that might jeopardise the structural integrity of the plasma facing components. It is, therefore, of paramount importance to have appropriate tools to reproduce the thermal-hydraulic behaviour of tokamak cooling systems during major operational and accidental scenarios in a realistic and reliable way. In this context, University of Palermo in cooperation with EUROfusion has developed a finite volume model of the Primary Heat Transport System (PHTS) feeding the EU-DEMO Water Cooled Lithium Lead Breeding Blanket (WCLL BB). The activity has been led following a theoretical–computational approach based on the adoption of the TRACE thermal-hydraulic system code. Particular attention has been paid to capturing all the main geometrical, hydraulic and heat transfer features characterising both in-vessel and ex-vessel components. Preliminary analyses have also been carried out to check the code’s predictive potential in fusion relevant applications. Models, assumptions, and outcomes of the analyses are herewith reported and critically discussed.

Design Features and Simulation of the New-Build HELOKA-US Facility for the Validation of the DEMO Helium-Cooled Pebble Bed Intermediate Heat Transport and Storage System

For the EU-DEMO Helium-Cooled Pebble Bed (HCPB) concept, an indirect coupled design (ICD) with a molten salt (MS) loop as an intermediate heat transport and storage system (IHTS) is considered for the conceptual design phase. The IHTS with an energy storage decouples the primary heat transport system (PHTS) that undergoes pulse and dwell power cycles from the power conversion system (PCS), and thus can provide stable power to the turbine and grid. However, the maintenance of stable He and MS parameters during transitions from dwell to pulse and vice versa is challenging for the design of the MS loop, and the real performance of the helium–MS heat exchanger (He/MS HX) shall be verified. To investigate such components and conditions, a new R&D infrastructure HELOKA-US (Helium Loop Karlsruhe—Upgrade Storage) is under construction for the validation of prototypical components and the MS loop operation under stationary and transitional conditions. This paper provides the design features of Phase 1a of the project and the simulation results with EBSILON on the power generation phase.

Interaction of magnesium ions in high temperature aqueous solution with Incoloy-800

Corrosion and activated corrosion product deposition in the out of core surfaces is a major concern in Pressurized Heavy Water Reactors (PHWRs) as it causes radiation exposure to plant personnel during routine maintenance. The ability of added Mg 2+ ions in the coolant to mitigate these undesirable effects in Carbon Steel structural material is known. By contrast, its effect on other heat transfer surfaces in the Primary Heat Transport (PHT) system is largely unknown. This paper presents a detailed study on the role of Mg 2+ ions in modifying the oxide on Incoloy-800.Incoloy-800 was exposed to normal and Mg 2+ ion modified alkaline water chemistry in a once through loop. Mg 2+ incorporation into the ferrite lattice was indicated in Raman, GIXRD analysis and corroborated by XPS technique. The incorporation of Mg 2+ ions in the oxide films resulted in a higher total impedance and reduced passive current density when compared to the films without Mg 2+ ions. Thus oxide with Mg 2+ ions improves the general corrosion resistance and does not induce localised corrosion in the material. • Passive films formed on Incoloy under simulated PHT conditions, incorporated Mg 2+ ions from solution into the ferrite lattice. • Mg Modified films were more compact and less porous. • Higher impedance and lower point defects were observed for Mg containing films. • Mg incorporation resulted in reduction of metal release at metal/oxide interface.

A Critical Review of Radiolysis Issues in Water-Cooled Fission and Fusion Reactors: Part II, Prediction of Corrosion Damage in Operating Reactors

The radiolysis of water is a significant cause of corrosion damage in the primary heat transport systems (PHTSs) of water-cooled, fission nuclear power reactors (BWRs, PWRs, and CANDUs) and is projected to be a significant factor in the evolution of corrosion damage in future fusion reactors (e.g., the ITER that is currently under development). In Part I of this two-part series, we reviewed the proposed mechanisms for the radiolysis of water and demonstrate that radiolysis leads to the formation of a myriad of oxidizing and reducing species. In this Part II, we review the role that the radiolysis species play in establishing the electrochemical corrosion potential (ECP) and the development of corrosion damage due to intergranular stress corrosion cracking (IGSCC) in reactor PHTSs. We demonstrate, that the radiolytic oxidizing radiolysis products, such as O2, H2O2, HO2−, and OH, when in molar excess over reducing species (H2, H, and O22−), some of which (H2) are preferentially stripped from the coolant upon boiling in a BWR PHTS, for example, renders the coolant in many BWRs oxidizing, thereby shifting the ECP in the positive direction to a value that is more positive than the critical potential (Ecrit = −0.23 Vshe at 288 °C) for IGSCC in sensitized austenitic stainless steel (e.g., Type 304 SS). This has led to many IGSCC incidents in operating BWRs over the past five decades that has exacted a great cost on the plant operators and electricity consumers, alike. In the case of PWRs, the primary circuits are pressurized with hydrogen to give a hydrogen concentration of 10 to 50 cm3/kgH2O (0.89 to 4.46 ppm), such that no sustained boiling occurs, and the hydrogen suppresses the radiolysis of water, thereby inhibiting the formation of oxidizing radiolysis products from water. Thus, the ECP is dominated by the hydrogen electrode reaction (HER), although important deviations from the HER equilibrium potential may occur, particularly at low [H2]. In any event, the ECP is displaced to approximately −0.85 Vshe, which is below the critical potential for IGSCC in sensitized stainless steels but is also more negative than the critical potential for the hydrogen-induced cracking (HIC) of mill-annealed Alloy 600. This has led to extensive cracking of steam generator tubing and other components (e.g., control rod drive tubes, pressurizer components) in PWRs that has also exacted a high cost on operators and power consumers. Although the ITER has yet to operate, the proposed chemistry protocol for the coolant places it close to a BWR operating on Normal Water Chemistry (NWC) without boiling or, if hydrogen is added to the IBED-PHTS, close to a BWR on Hydrogen Water Chemistry (HWC). In the current ITER technology, the concentration of H2 in the IBED-PHTS is specified to be 80 ppb, which is the concentration that will be experienced in both the Plasma Flux Area (PFA) and in the Out of Plasma Flux Area (OPFA). That corresponds to 0.90 cc(STP) H2/KgH2O, compared with 20–50 cc(STP) H2/KgH2O employed in a PWR primary coolant circuit and 5.5 to 22 cc(STP) H2/KgH2O in a BWR on hydrogen water chemistry (HWC). We predict that a hydrogen concentration of 80 ppb is sufficient to reduce the ECP in the OPFA to a level (−0.324 Vshe) that is sufficient to suppress the crack growth rate (CGR) below the practical, maximum level of 10−9 cm/s (0.315 mm/a) at which SCC is considered not to be a problem in a coolant circuit but, in the PFA, the ECP is predicted to be 0.380 Vshe, which gives a calculated standard CGR of 2.7 × 10−6 cm/s. This is more than three orders in magnitude greater that the desired maximum value of 10−9 cm/s. We recommend that the HWC issue in ITER be revisited to develop a protocol that is effective in suppressing both the ECP and the CGR in the PFA to levels that permit the operation of the IBED-PHTS in accordance with the experience gained in fission reactor technology.

Neutronics and thermal-hydraulics coupling analyses on transient and accident behaviors of molten chloride salt fast reactor

ABSTRACT Neutronics and thermal-hydraulics coupling transient analyses have been carried out to investigate the intrinsic safety characteristics and fuel temperature ranges of a molten chloride salt fast reactor. The analysis system includes a primary heat transport system and a secondary heat transport system connected by aheat exchanger and a decay heat removal system connected to the secondary system. The neutronics and thermal-hydraulics coupling analysis was performed with the RELAP5-3D code to analyze the entire plant behavior of each event, the temperature and velocity of the reactor inlet flow were obtained, and the reactor power was analyzed in detail following the FLUENT code. Through these analyses, the most severe events for the reactor were revealed. Analysis using the two codes showed that the reactor power calculated by RELAP5-3D was in good agreement with that of FLUENT although the detailed flow inside the core could not be reproduced.

Modeling of Safety Basis Events in the VTR

The Versatile Test Reactor (VTR) is a fast spectrum test reactor currently being developed in the United States under the direction of the U.S. Department of Energy, Office of Nuclear Energy. Safety analysis of the conceptual VTR design is being performed using the SAS4A/SASSYS-1 fast reactor safety analysis code with a model representing the reactor core, primary and secondary heat transport systems, reactor vessel auxiliary cooling system, and reactor protection system. The system’s response and safety performance are being evaluated for a wide spectrum of event initiators and accident sequences. This paper presents an overview of the activities that are ongoing in support of the modeling and analysis of safety basis events (SBEs) in the VTR, including the VTR SAS4A/SASSYS-1 model development, an overview of the SAS4A/SASSYS-1 verification and validation efforts, and a summary of key model development activities to improve the predictive capability of the code. A summary of the results and an analysis of several key SBEs are also presented. VTR authorization from the U.S. Department of Energy will require transient simulations that are demonstrated to be accurate.

Reliability-based inspection and maintenance planning of a nuclear feeder piping system

Inspection and maintenance (I&M) is essential to ensure the integrity of feeder pipes, which are parts of the primary heat transport system in a nuclear power plant. The pipes are subject to flow accelerated corrosion (FAC), which can compromise the integrity of the piping system and lead to high repair costs. We explore the opportunity for improving I&M strategies while ensuring that the system still maintains an acceptable level of reliability. To this aim, a reliability-based planning framework is proposed, in which every pipe in the system meets the minimum thickness requirement at a specified annual probability. With this planning framework we can a) evaluate the performance of any I&M strategy constrained to a fixed reliability criterion, without requiring this strategy to be specifically designed for such a criterion; and b) find an I&M strategy optimized for this reliability level using a heuristic description of the strategy space. We demonstrate the framework with a case study, where the wall thinning due to FAC is modeled as a Gamma process with uncertain parameters. We compare the expected life-cycle cost of multiple strategies for I&M of a feeder system with 480 pipes. The proposed approach is compared with an I&M strategy currently used by the industry, which highlights the efficiency of the proposed optimization method.

Preliminary Analysis of Unprotected Transients in the VTR

The Versatile Test Reactor (VTR) is a fast spectrum test reactor currently being developed in the United States under the direction of the U.S. Department of Energy, Office of Nuclear Energy. Safety analysis of the conceptual VTR design is being performed using the SAS4A/SASSYS-1 fast reactor safety analysis code with a model representing the reactor core, primary and intermediate heat transport systems, reactor vessel auxiliary cooling system, and reactor protection system. The system’s response and safety performance has been evaluated for a wide spectrum of event initiators and accident sequences. This paper focuses on the results for several unprotected transient scenarios where the reactor protection system is assumed to fail to take any action. Even without the reactor protection system, the strong reactivity feedback response of the core reduces power to safe levels matching the available heat rejection. In the station blackout transient, the primary heat transport system is able to transition quickly and effectively to natural circulation. At the current stage of design, transient simulation results for the VTR indicate that large safety margins exist for many event initiators, including the unprotected transients presented in this paper.

Effects of zinc-aluminium injection on corrosion behaviour and semiconductor properties of carbon steel in the PHT system of PHWR.

Carbon steel is widely used as the piping of the primary heat transport system (PHT) in pressurized heavy water reactors (PHWR). Effects of zinc/aluminum treatment and simultaneous injection of zinc and aluminum on corrosion behavior and semiconductor properties of oxide films formed on carbon steels were characterized by the gravimetric method, potentiodynamic polarization scan, EIS, Mott–Schottky test, SEM, EDS mapping, XPS analysis and photocurrent response measurement. The results showed that all the metal cation ions treatment can improve the corrosion resistance of oxide films in varying degrees. 20 ppb Zn2+ had the greatest enhancement in corrosion resistance, followed by 20 ppb Zn2+ + 20 ppb Al3+. ZnFe2O4, FeAl2O4 and ZnAl2O4 were detected to be new spinel phases generated in oxide films. The oxide films on the surface of carbon steel all demonstrated n-type semiconductor properties. It was worth noting that the total content of manganese and zinc in the oxide film played an important role in the corrosion resistance of carbon steel.

Estimating degradation growth rate and time of component replacement from limited inspection data using mixed-effects modelling

Predicting the progress of degradation of a specific engineering component can be highly uncertain, due to a lack of repeat observations over time. This paper illustrates how the mixed-effects modelling approach can be used to significantly reduce the uncertainty in the component specific predictions by leveraging the available information from the larger component population. The methodology is applied to the assessment of flow accelerated corrosion (FAC) of a large population of piping in the primary heat transport system of a nuclear reactor. The pipe wall thickness loss is modelled using the linear mixed-effects regression (LMER) approach, and applied at the group level based on pipe geometry. The results demonstrate how the mixed-effects approach provides a highly effective and consistent method for combining all of the valuable (and often highly expensive) inspection data into a unified framework to obtain more robust and credible estimates of degradation growth rates and times of replacement of all components in the population, including uninspected components and components with very limited data (e.g., one or two inspections).

Modification of oxide film on Zircaloy-2 by magnesium ions in high temperature aqueous solution

Inorganic metal ion additives are being explored for controlling the corrosion and deposition of activated corrosion products on out of core surfaces in Pressurized Heavy Water Reactors (PHWRs). Mg2+ ions can modify the oxides on heat transfer surfaces like Zircaloy and thus influence its corrosion properties. Zircaloy-2 coupons were exposed to simulated Primary Heat Transport (PHT) system environment in the presence and absence of Mg2+ ions and were characterized to understand the surface morphology, defect density and the corrosion protectiveness. Incorporation of Mg into the oxide was confirmed by XPS. GIXRD and Raman spectra showed presence of predominantly monoclinic zirconia with traces of tetragonal phase. GIXRD on Mg-doped films showed that, the monoclinic ZrO2 fraction increased by 2% from 7d to 42d exposure. The films formed in presence of Mg always had higher impedance and lower passive current density than that formed in the absence of Mg for the same duration. This study shows that the Mg doped oxide imparted corrosion protectiveness and the doping is not detrimental to the material.

Scaling Methodology for Integral Effects Tests in Support of Fluoride Salt–Cooled High-Temperature Reactor Technology

The capability to validate integral transient response models is of critical importance for licensing new reactor designs. The Kairos Power testing program has developed a methodology to design scaled experiments that predict the thermal fluid behavior of the Kairos Power Fluoride Salt–Cooled High-Temperature Reactor (KP-FHR). Such experiments will be used as part of the assessment base of evaluation models supporting KP-FHR safety analysis. The Hierarchical Two-Tiered Scaling (H2TS) methodology was selected for Kairos Power scaling efforts that will be applied to integral effects tests (IETs) for system-level testing. In this paper, the scaling methodology is presented for thermal fluid IETs that will model the KP-FHR primary heat transport system under normal operations and transient conditions involving transition to natural circulation. This paper provides a basis for using surrogate fluids for testing that requires the thermal fluid performance of the KP-FHR primary coolant lithium fluoride–beryllium fluoride [2LiF/BeF2 (Flibe)] to be replicated. Kairos Power intends to use heat transfer oil as a surrogate fluid for Flibe in thermal fluid IETs. This paper demonstrates that this class of surrogate fluids is an acceptable substitute for Flibe salt for some types of scaled IETs and that the principal thermal fluid properties can be properly scaled with minor distortions over the range of conditions expected for both normal and off-normal operating conditions of the KP-FHR.

Improvement of aseismic performance of a PGSFR PHTS pump

A design study was performed to improve the limit aseismic performance (LSP) of a primary heat transport system (PHTS) pump. This pump is part of the primary equipment of a prototype generation IV sodium-cooled fast reactor (PGSFR). The LSP is the maximum allowable seismic load that still ensures structural integrity. To calculate the LSP of the PHTS pump, a structural analysis model of the pump was developed and its dynamic characteristics were obtained by modal analysis. The floor response spectrum (FRS) initiated from a safety shutdown earthquake (SSE), 0.3 g, was applied to the support points of the PHTS pump, and then the seismic induced stresses were calculated. The structural integrity was evaluated according to the ASME code, and the LSP of the PHTS pump was calculated from the evaluation results. Based on the results of the modal analysis and LSP of the PHTS pump, design parameters affecting the LSP were selected. Then, ways to improve the LSP were proposed from sensitivity analysis of the selected design variables.