Achievements and further plans for the OECD Halden reactor project materials programme

The materials programme at Halden, in addition to cladding corrosion studies, is aimed also at addressing the effects of operating conditions and water chemistry variables on core materials behaviour, particularly as related to reactor pressure vessel integrity and irradiation assisted stress corrosion cracking (IASCC), the materials degradation phenomenon which affects the structural integrity of stainless steel and nickel based components. The aim of the experimental work is to improve the understanding of materials ageing processes, to demonstrate the benefits of mitigation measures and to evaluate properties of materials, which have been subjected to long in-reactor service. While a number of the studies are performed in loops, which simulate light water reactor environments in terms of thermal-hydraulic, radiation and water chemistry conditions, dry irradiation facilities are also utilised, particularly in relation to studies aimed at determining the effects of fluence on material integrity.     

Implementation of an integrated on-line process surveillance and diagnostic system at the Halden reactor project: MOAS

MOAS is an integrated on-line process surveillance and diagnostic system that uses several different models for knowledge acquisition and diagnostic reasoning, such as goal-tree success-tree model, process monitor trees, and sensor failure diagnosis trees. Within these models, the knowledge of the process and its operation, including deep knowledge, like mass balance or controller algorithm, is incorporated. During an extensive review, made as part of the integrated diagnosis system project of the Halden reactor project, MOAS (Maryland Operator Advisory System) was identified as one of the most thorough systems developed thus far. MOAS encompasses diverse functional aspects that are required for an effective process disturbance management: (1) intelligent process monitoring and alarming, (2) on-line sensor data validation and sensor failure diagnosis, (3) on-line hardware (besides sensors) failure diagnosis, and (4) real-time corrective measure synthesis. The MOAS methodology was used for the NORS (Nokia Research Simulator) process at the Halden man-machine laboratory HAMMLAB of the OECD Halden reactor project. The performance tests of MOAS, implemented in G2 real-time expert system shell, show that MOAS successfully carries out its intended functions, i.e. quickly recognizing an occurring disturbance, correctly diagnosing its cause, and presenting advice on its control to the operator. The lessons learned and insights gained during the implementation and performance tests also are discussed.     

The chemistry and thermodynamics of molten salt reactor fuels

A fairly detailed knowledge of the chemistry and thermodynamics of molten salt breeder reactor fuels, typically LiF-BeF₂ThF ₄-UF₄ ($\text{72-16-12- < 1}\text{mol}\text{\%}$), has been gained from measurements of heterogeneous equilibria involving various gases (e.g. HF, H₂O) and solids (usually oxides). This information combined with the thermochemical data available for the pure compounds has given formation free energies for the fual components. Their entropies can be estimated from the known entropies of the corresponding solid oxides and the charge on the cation. Their activity coefficients depend primarily on the mol % of LiF in the salt mixture. The solubility of oxides generally decreases with the square of the cation charge divided by the radius. Thus the actinide dioxides and, especially, Pa₂O₂ (or an addition compound of it) are of low solubility. Pourbaix diagrams are presented which summarize the conditions of redox potential and oxide concentration wherein the fuel may be contained without significant corrosion or precipitation reactions, or conditions wherein selective oxide precipitation may be carried out for purposes of fuel reprocessing.     

The DEEPSSI project, design, testing and modeling of steam injectors

The DEEPSSI project is a steam injector research programme. Among thermal-hydraulic passive systems, the steam injectors (also called “condensing ejectors” or “steam jet pumps”) are very interesting apparatus with very specific characteristics (high velocity, very low pressure). The envisaged reactor application is the Steam Generator Emergency FeedWater System (EFWS) of Pressurised Water Reactors (PWRs). The heart of this project is the development and the testing of an innovative steam injector design. Three experimental facilities are involved: CLAUDIA in France, IETI in Italy and IMP-PAN in Poland. In these facilities, different design options have been tested and some significant improvements of the initial design have been obtained.In addition to the experimental studies, the development of a steam injector computational model has been undertaken in order to model industrial systems based on steam injectors. The one-dimensional module of the system code CATHARE2 has been chosen to be the basis of this model. The first results obtained have confirmed the capabilities of CATHARE2 to describe the steam injector thermal-hydraulics.     

BNCT project at Tehran Research Reactor: Current and prospective plans

Boron Neutron Capture Therapy (BNCT) for brain tumor treatment is under development at the Tehran Research Reactor (TRR). This paper presents all current research activities that were performed during recent years as well as the prospective of BNCT research at TRR. The theoretical and experimental investigations show that TRR has a very good potential to consider it as a pilot facility for BNCT research in the Middle East and could be facilitated for clinical applications. In this way, there are some steps and also some challenges which are described in the paper.     

SNR 300 fast breeder reactor: steel containment - design, erection, testing

The Kalkar Nuclear Power Plant which is equipped with an 300 MW fast breeder reactor is being built by a Consortium mainly comprising German, Belgian and Dutch companies.The components of the fast breeder reactor are enclosed in a concrete containment which is designed to withstand severe external and internal loading.The concrete enclosure is surrounded by a steel containment which is designed to prevent the release of radioactivity following a postulated accident involving the nuclear components inside the concrete containment.The paper describes the solutions adopted for the different parts of the steel containment, the calculations verifying the suitability of the designs, the erection and the steel containment pressure and leak tests. The tests were performed with successful results in 1984.     

A US perspective on fast reactor fuel fabrication technology and experience. Part II: Ceramic fuels

This paper is Part II of a review focusing on the United States experience with oxide, carbide, and nitride fast reactor fuel fabrication. Over 60 years of research in fuel fabrication by government, national laboratories, industry, and academia has culminated in a foundation of research and resulted in significant improvements to the technologies employed to fabricate these fuel types. This part of the review documents the current state of fuel fabrication technologies in the United States for each of these fuel types, some of the challenges faced by previous researchers, and how these were overcome. Knowledge gained from reviewing previous investigations will aid both researchers and policy makers in forming future decisions relating to nuclear fuel fabrication technologies.     

Acoustic emission for on-line reactor monitoring: Results of intermediate vessel test monitoring and reactor hot functional testing

This article discusses a program designed to develop the use of acoustic emission (AE) methods for continuous surveillance to detect and evaluate flaw growth in reactor pressure boundaries. Technology developed in the laboratory for identifying AE from crack growth and for using that AE information to estimate flaw severity is now being evaluated on an intermediate vessel test and on a reactor facility.A vessel, designated ZB-1, has been tested under fatigue loading with simulated reactor conditions at Mannheim, West Germany, in collaboration with the German Materialprüfungsanstalt (MPA), Stuttgart. Fatigue cracking from machined flaws and in a fabrication weld were both detected clearly by AE.AE data were measured on a US nuclear reactor (Watts Bar, Unit 1) during hot functional preservice testing. This demonstrated that coolant flow noise is a manageable problem and that AE can be detected under operational coolant flow and temperature conditions.     

A US perspective on fast reactor fuel fabrication technology and experience part I: metal fuels and assembly design

This paper is part I of a review focusing on the United States experience with metallic fast reactor fuel fabrication and assembly design for the Experimental Breeder Reactor-II (EBR-II) and the Fast Flux Test Facility (FFTF). Experience with metal fuel fabrication in the United States is extensive, including over 60 years of research conducted by the government, national laboratories, industry, and academia. This experience has culminated in a considerable amount of research that resulted in significant improvements to the technologies employed to fabricate metallic fast reactor fuel. This part of the review documents the current state of fuel fabrication technologies for metallic fuels, some of the challenges faced by previous researchers, and how these were overcome. Knowledge gained from reviewing previous investigations will aid both researchers and policy makers in forming future decisions relating to nuclear fuel fabrication technologies.     

The Fort St. Vrain high temperature gas-cooled reactor: XII. The dew point moisture monitor testing program

Moisture ingress into the core volume could cause damaging reactions with the moderator-reflector graphite and burnable poison, therefore a dew point moisture monitoring system has been developed with the basic design criteria that a plant protective system trip is signaled after the system detects high primary coolant helium moisture levels and that the system is able to correctly identify which of two steam generator loops is leaking. Modifications to the sample supplies to the monitors were necessary to reduce the system's unsatisfactory response time at lower reactor power levels.     

The EU TBM systems: Design and development programme

This paper presents the status of the design and of the development programme of the two test blanket systems (TBSs) based on the blanket concepts supported by the EU, namely the helium cooled lithium lead (HCLL) and helium cooled pebble bed (HCPB) concepts.Both the test blanket modules (TBMs) box design and the associated systems (Helium Cooling Systems, PbLi loop for the HCLL system, helium processing systems for tritium extraction, etc.) have been revised and, where needed, modified according to the assumption that one ITER equatorial port could be available for testing the two European test blanket modules (TBMs).According to EU TBMs programme, two reliable test blanket systems shall be ready for installation on the first day of ITER operation. In order to comply with this ambitious objective, six EURATOM associates who have sustained the TBM program so far have joined themselves in a consortium aiming to ensure an efficient management of the project tasks and exploit specific competences enhancing potential synergies. The consortium objectives and development programme are summarised in the paper.     

The experimental breeder reactor II inherent shutdown and heat removal tests — results and analysis

A test program is being conducted to demonstrate that a power-producing liquid-metal reactor (LMR) can (1) passively remove shutdown heat by natural convection, (2) passively reduce power in response to a loss of reactor flow, and (3) passively reduce power in response to a loss of the balance-of-plant heat sink. Measurements and pretest predictions confirm that natural convection is a reliable, predictable method of shutdown heat removal and suggest that safety-related pumps or pony motors are not necessary for safe shutdown heat removal in an LMR. Measurements from tests in which reactor flow and heat rejection to the balance of plant were perturbed show that reactivity feedbacks can passively control power and temperature. Data from these tests form a basis for additional tests including a complete loss of flow without scram and a complete loss of heat sink without scram.     

The advanced high-temperature reactor: High-temperature fuel, liquid salt coolant, liquid-metal-reactor plant

The Advanced High-Temperature Reactor is a new reactor concept that combines four existing technologies in a new way: (1) coated-particle graphite-matrix nuclear fuels (traditionally used for helium-cooled reactors), (2) Brayton power cycles, (3) passive safety systems and plant designs from liquid-metal-cooled fast reactors, and (4) low-pressure liquid-salt coolants with boiling points far above the maximum coolant temperature. The new combination of technologies enables the design of a large [2400- to 4000-MW(t)] high-temperature reactor, with reactor-coolant exit temperatures between 700 and 1000°C (depending upon goals) and passive safety systems for economic production of electricity or hydrogen. The AHTR [2400-MW(t)] capital costs have been estimated to be 49 to 61% per kilowatt (electric) relative to modular gas-cooled [600-MW(t)] and modular liquid-metal-cooled reactors [1000-MW(t)], assuming a single AHTR and multiple modular units with the same total electrical output. Because of the similar fuel, core design, and power cycles, about 70% of the required research is shared with that for high-temperature gas-cooled reactors.     

Neutron embrittlement of VVER reactor pressure vessels—recent results, open issues and new developments

Decisions regarding the verification of design plant lifetime involve a determination of the component and circuit condition. Neutron embrittlement of reactor pressure vessels (RPV) becomes a crucial consideration for continued safe plant operation. Since 1991, the European Commission (EC) has financed a significant number of projects in this area, in particular through the TACIS and PHARE programmes. In Nuclear Safety support programmes the countries mainly concerned are Russia, Ukraine, Armenia, and Kazakhstan for the TACIS programme, and Bulgaria, Czech Republic, Hungary, Slovak Republic, Lithuania, Romania and Slovenia for the PHARE programme. The corresponding results of these projects on neutron embrittlement have contributed to a better understanding of the available data and their uncertainties. Recent confirmations of the irradiation temperatures of the surveillance specimens in the operating VVER nuclear power plants show that the surveillance specimens can be used further for the validation of the current and the expected neutron embrittlement. Taking into account the latest results as a baseline for the development of the necessary material data to be used for upgraded RPV integrity assessment, two new TACIS projects are being launched, jointly with Russia and Ukraine, whose scope is:

• Upgrading the surveillance databases.

• Acquisition of new impact test and toughness results on reconstituted surveillance specimens, including the evaluation of the Mater Curve Approach.

• Further validation tests of the shape of the fracture toughness curve and the base and weld metal and characterization of the cladding.

• Preparation of some selected upgraded RPV integrity assessments, with insights on the latest approved methodology.

The paper summarizes the major conclusions of the recent completed EC projects, reviews the remaining major open issues in the field of reliable determination of fracture toughness properties of the operating VVER RPVs, and details the scope of the new projects.     

Nuclear energy for transportation: Paths through electricity, hydrogen and liquid fuels

The transportation sector consumes about a quarter of final energy in Japan and worldwide, and presently most of the energy is supplied by petroleum. For global environment and resources, it is important to seek possibilities of replacing a substantial part of the transportation energy by nuclear energy.

For supplying nuclear energy to the transportation sector, investigated are the paths through such ‘energy carriers’ as electricity, hydrogen and synthetic liquid fuels, to the means of transportation such as automobiles. These energy carriers can be produced from nuclear energy, by itself or synergistically with other primary energies like fossil fuels or biomass.

In this paper, possibilities and impacts of these energy carriers to power transportation means are reviewed, and measures and tasks to supply these energy carriers from nuclear energy are examined.

In converting the primary energies into the energy carriers, the synergistic process may be more advantageous than the individual process. Some of the exploratory processes to produce synthetic liquid fuels from fossil fuels and nuclear energy are presented.     

PARIS project: Radiolytic oxidation of molecular iodine in containment during a nuclear reactor severe accident: Part 2. Formation and destruction of iodine oxides compounds under irradiation - Experimental results modelling

In the case of a severe accident in a nuclear Light Water Reactor (LWR), the high radiation fields reached in the reactor containment building due to the release of fission products from the reactor core would induce air radiolysis. The air radiolysis products (ARP) could, in turn, oxidise gaseous molecular iodine (I₂) into aerosol-borne iodine-oxygen-nitrogen compounds, abbreviated as iodine oxides (IO_x). These reactions involve the conversion of a gaseous iodine compound resulting in a change of the iodine depletion rate from the containment atmosphere. Kinetic data were produced within the first part of PARIS project on the air radiolysis products formation and destruction. The second part of the PARIS project as presented in this paper deals with the impact of the ARP on the conversion of I₂ into IO_x. The objective was to provide a database to develop new or to validate existing kinetic models of formation and destruction of iodine oxides.The iodine tests of the PARIS project, performed at very low, realistic iodine concentrations, constitute an important database to further develop or validate empirical and mechanistic models on radiolytic I₂ oxidation. In the presence of painted surface areas or silver aerosol surface areas, radiolytic I₂ oxidation is negligible compared to I₂ adsorption on these surfaces for the conditions examined. However, radiolytic I₂ oxidation remains very efficient if surface areas are small or if they are made of the relatively non-reactive stainless steel.     

The nuclear safety project at KfK 1972–1986 — Objectives and results

In 1972 the light water reactor safety activities conducted at the Karlsruhe Nuclear Research Center (KfK) were combined under the Nuclear Safety Project (PNS). Its primary objective was to assess in quantifiable terms the safety reserves which are provided in nuclear power plant design in a conservative approach. While in the initial phase R&D work conducted under the project was largely characterized by investigations of the design basis accidents, mainly the loss-of-coolant accident, emphasis in the past decade has been shifted more and more towards severe core and core meltdown accident analysis. The activities comprise both theoretical studies and experimental investigations, often performed in adequate, large-scale facilities. All activities have been an essential part of the reactor safety research program of the Federal Ministry for Research and Technology (BMFT) and have been coordinated with a number of other programs conducted in Germany and abroad. This paper gives a broad overview of PNS contributions to LWR safety research in the past 15 years and summarizes the results, comparing them with the general goals defined. In conclusion, the attempt is made to give an outlook on remaining activities in LWR safety research being carried out by KfK.