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DOE OSTI2012

An extended conventional fuel cycle for the B and W mPower{sup TM} small modular nuclear reactor

The B and W mPower{sup TM} reactor is a small pressurized water reactor (PWR) with an integral once-through steam generator and a thermal output of about 500 MW; it is intended to replace aging fossil power plants of similar output. The core is composed of 69 reduced-height PWR assemblies with the familiar 17 x 17 fuel rod array. The Babcock and Wilcox Company (B and W) is offering a core loading and cycle management plan for a four-year cycle based on its presumed attractiveness to potential customers. This option is a once-through fuel cycle in which the entire core is discharged and replaced after four years. In addition, a conventional fuel utilization strategy, employing a periodic partial reload and shuffle, was developed as an alternative to the four-year once-through fuel cycle. This study, which was performed using the Studsvik core design code suite, is a typical multi-cycle projection analysis of the type performed by most fuel management organizations such as fuel vendors and utilities. In the industry, the results of such projections are used by the financial arms of these organizations to assist in making long-term decisions. In the case of the B and W mPower reactor, this analysis demonstrates flexibility for customers who consider the once-through fuel cycle unacceptable from a fuel utilization standpoint. As expected, when compared to the once-through concept, reloads of the B and W mPower reactor will achieve higher batch average discharge exposure, will have adequate shut-down margin, and will have a relatively flat hot excess reactivity trend at the expense of slightly increased peaking. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Evaluation of the coolant reactivity coefficient influence on the dynamic response of a small LFR system

An assessment of the coolant reactivity feedback influence on a small Lead-cooled Fast Reactor (LFR) dynamics has been made aimed at providing both qualitative and quantitative insights into the system transient behavior depending on the sign of the above mentioned coefficient. The need of such an investigation has been recognized since fast reactors cooled by heavy liquid metals show to be characterized by a strong coupling between primary and secondary systems. In particular, the coolant density and radial expansion coefficients have been attested to play a major role in determining the core response to any perturbed condition on the Steam Generator (SG) side. The European Lead-cooled System (ELSY)-based demonstrator (DEMO) has been assumed as the reference LFR case study. As a first step, a zero-dimensional dynamics model has been developed and implemented in MATLAB/SIMULINK{sup R} environment; then typical transient scenarios have been simulated by incorporating the actual negative lead density reactivity coefficient and its opposite. In all the examined cases results have shown that the reactor behaves in a completely different way when considering a positive coolant feedback instead of the reference one, the system free dynamics resulting moreover considerably slower due to the core and SG mutually conflicting reactions. The outcomes of the present analysis may represent a useful feedback for both the core and the control system designers. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

ESBWR response to an extended station blackout/loss of all AC power

U.S. federal regulations require light water cooled nuclear power plants to cope with Station Blackouts for a predetermined amount of time based on design factors for the plant. U.S. regulations define Station Blackout (SBO) as a loss of the offsite electric power system concurrent with turbine trip and unavailability of the onsite emergency AC power system. According to U.S. regulations, typically the coping period for an SBO is 4 hours and can be as long as 16 hours for currently operating BWR plants. Being able to cope with an SBO and loss of all AC power is required by international regulators as well. The U.S. licensing basis for the ESBWR is a coping period of 72 hours for an SBO based on U.S. NRC requirements for passive safety plants. In the event of an extended SBO (viz., greater than 72 hours), the ESBWR response shows that the design is able to cope with the event for at least 7 days without AC electrical power or operator action. ESBWR is a Generation III+ reactor design with an array of passive safety systems. The ESBWR primary success path for mitigation of an SBO event is the Isolation Condenser System (ICS). The ICS is a passive, closed loop, safety system that initiates automatically on a loss of power. Upon Station Blackout or loss of all AC power, the ICS begins removing decay heat from the Reactor Pressure Vessel (RPV) by (i) condensing the steam into water in heat exchangers located in pools of water above the containment, and (ii) transferring the decay heat to the atmosphere. The condensed water is then returned by gravity to cool the reactor again. The ICS alone is capable of maintaining the ESBWR in a safe shutdown condition after an SBO for an extended period. The fuel remains covered throughout the SBO event. The ICS is able to remove decay heat from the RPV for at least 7 days and maintains the reactor in a safe shutdown condition. The water level in the RPV remains well above the top of active fuel for the duration of the SBO event. Beyond 7 days, only a few simple actions are needed to cope with the SBO for an indefinite amount of time. The operation of the ICS as the primary success path for mitigation of an SBO, allows for near immediate plant restart once power is restored. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Olkiluoto 1 and 2 - Plant efficiency improvement and lifetime extension-project (PELE) implemented during outages 2010 and 2011

Teollisuuden Voima Oyj (TVO) is a non-listed public company founded in 1969 to produce electricity for its stakeholders. TVO is the operator of the Olkiluoto nuclear power plant. TVO follows the principle of continuous improvement in the operation and maintenance of the Olkiluoto plant units. The PELE project (Plant Efficiency Improvement and Lifetime Extension), mainly completed during the annual outages in 2010 and 2011, and forms one part of the systematic development of Olkiluoto units. TVO maintains a long-term development program that aims at systematically modernizing the plant unit systems and equipment based on the latest technology. According to the program, the Olkiluoto 1 and Olkiluoto 2 plant units are constantly renovated with the intention of keeping them safe and reliable, The aim of the modernization projects is to improve the safety, reliability, and performance of the plant units. PELE project at Olkiluoto 1 was done in 2010 and at Olkiluoto 2 in 2011. The outage length of Olkiluoto 1 was 26 d 12 h 4 min and Olkiluoto 2 outage length was 28 d 23 h 46 min. (Normal service-outage is about 14 days including refueling and refueling-outage length is about seven days. See figure 1) The PELE project consisted of several single projects collected into one for coordinated project management. Some of the main projects were as follows: - Low pressure turbines: rotor, stator vane, casing and turbine instrumentation replacement. - Replacement of Condenser Cooling Water (later called seawater pumps) pumps - Replacement of inner isolation valves on the main steam lines. - Generator and the generator cooling system replacement. - Low voltage switchgear replacement. This project will continue during future outages. PELE was a success. 100 TVO employees and 1500 subcontractor employees participated in the project. The execution of the PELE projects went extremely well during the outages. The replacement of the low pressure turbines and seawater pumps improved the efficiency of the plant units, and a power increase of nearly 20 MW was achieved at both plant units. PELE wonderfully manifests one of the strategic goals of our company; developing the competence of our in-house personnel by working in projects. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Development of an object-oriented dynamics simulator for a LFR DEMO

A control-oriented dynamics simulator for a Generation IV Lead-cooled Fast Reactor (LFR) demonstrator (DEMO) has been developed aimed at providing a flexible, simple and fast-running tool allowing to perform design-basis transient and stability analyses, and to lay the foundations for the study of the system control strategy. For such purposes, a model representing a compromise between accuracy and straightforwardness has been necessarily sought, and in this view an object-oriented approach based on the Modelica language has been adopted. The reactor primary and secondary systems have been implemented by assembling both component models already available in a specific thermal-hydraulic library, and ad hoc developed nuclear component models suitably modified according to the specific DEMO configuration. The resulting overall plant simulator, incorporating also the balance of plant, consists in the following essential parts: core, integrated steam generator/primary pump block, cold and hot legs, primary coolant cold pool, turbine, heat sink, secondary coolant pump. Afterwards, the reactor response to typical transient initiators has been investigated: feedwater mass flow rate and temperature enhancement, turbine admission valve coefficient variation, increase of primary coolant mass flow rate, and transient of overpower have been simulated; results have been compared with the outcomes of analogous analyses performed by employing a lumped-parameter DEMO plant model. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Preliminary LOCA analysis of the westinghouse small modular reactor using the WCOBRA/TRAC-TF2 thermal-hydraulics code

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (> 225 MWe) integral pressurized water reactor (iPWR) with all primary components, including the steam generator and the pressurizer located inside the reactor vessel. The reactor core is based on a partial-height 17x17 fuel assembly design used in the AP1000{sup R} reactor core. The Westinghouse SMR utilizes passive safety systems and proven components from the AP1000 plant design with a compact containment that houses the integral reactor vessel and the passive safety systems. A preliminary loss of coolant accident (LOCA) analysis of the Westinghouse SMR has been performed using the WCOBRA/TRAC-TF2 code, simulating a transient caused by a double ended guillotine (DEG) break in the direct vessel injection (DVI) line. WCOBRA/TRAC-TF2 is a new generation Westinghouse LOCA thermal-hydraulics code evolving from the US NRC licensed WCOBRA/TRAC code. It is designed to simulate PWR LOCA events from the smallest break size to the largest break size (DEG cold leg). A significant number of fluid dynamics models and heat transfer models were developed or improved in WCOBRA/TRAC-TF2. A large number of separate effects and integral effects tests were performed for a rigorous code assessment and validation. WCOBRA/TRAC-TF2 was introduced into the Westinghouse SMR design phase to assist a quick and robust passive cooling system design and to identify thermal-hydraulic phenomena for the development of the SMR Phenomena Identification Ranking Table (PIRT). The LOCA analysis of the Westinghouse SMR demonstrates that the DEG DVI break LOCA is mitigated by the injection and venting from the Westinghouse SMR passive safety systems without core heat up, achieving long term core cooling. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Experimental results of the QUENCH-16 bundle test on air ingress

The out-of-pile bundle experiment QUENCH-16 on air ingress was conducted in the electrically heated 21-rod QUENCH facility at KIT in July 2011. It was performed in the frame of the EC supported LACOMECO program. The test scenario included the oxidation of the Zircaloy-4 claddings in air following a limited pre-oxidation in steam, and involved a long period of oxygen starvation to promote interaction with the nitrogen. The primary aim was to examine the influence of the formed oxide layer structure on bundle coolability and hydrogen release during the terminal flooding phase. QUENCH-16 was thus a companion test to the earlier air ingress experiment, QUENCH-10, which was performed with strongly pre-oxidized bundle. Unlike QUENCH-10, significant temperature escalation and intensive hydrogen release were observed during the reflood phase. Post-test investigations of bundle cross sections reveal residual nitride traces at various elevations. The external part of the oxide scale is of porous structure due to re-oxidation of nitrides during reflood. Relative thick internal oxide scales underneath this porous layer and residual nitrides were formed during reflood. At lower bundle elevations frozen partially oxidized melt was detected, relocated from upper elevations. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Coupled simulation of the reactor core using CUPID/MASTER

The CUPID is a component-scale thermal hydraulics code which is aimed for the analysis of transient two-phase flows in nuclear reactor components such as the reactor vessel, steam generator, containment. This code adopts a three-dimensional, transient, two-phase and three-field model, and includes physical models and correlations of the interfacial mass, momentum, and energy transfer for the closure. In the present paper, a multi-physics simulation was performed by coupling CUPID with a three dimensional neutron kinetics code, MASTER. MASTER is merged into CUPID as a dynamic link library (DLL). The APR1400 reactor core during a control rod drop/ejection accident was simulated as an example by adopting a porous media approach to employ a fuel assembly. The following sections present the numerical modeling for the reactor core, coupling of the kinetics code, and the simulation results. And also, a preliminary study for multi-scale simulation between CUPID and system-scaled thermal hydraulics code, MARS will be introduced as well. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Numerical study on interaction of local air cooler with stratified hydrogen cloud in a large vessel

Within the framework of the ERCOSAM project, planning calculations are performed to examine sensitivity parameters that can affect the break-up (erosion) of a helium layer by mitigation devices (i.e., cooler, spray, or Passive Autocatalytic Recombiner - PAR). This paper reports the GOTHIC analysis results for the cooler tests to be performed in the PANDA facility. The cooler elevation and geometry, helium layer thickness, steam distribution in the vessel, and the vessel geometry (inter-connected multi-compartments versus a single volume) on the erosion process as well as the cooling capacity are studied. This analysis is valuable because only a limited number of conditions will be examined in the planned experiments. The study provides a useful understanding of the interaction of a cooler with a stratified atmosphere. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Condensation in horizontal heat exchanger tubes

Many innovative reactor concepts for Generation III nuclear power plants use passive safety equipment for residual heat removal. These systems use two phase natural circulation. Heat transfer to the coolant results in a density difference providing the driving head for the required mass flow. By balancing the pressure drop the system finds its operational mode. Therefore the systems depend on a strong link between heat transfer and pressure drop determining the mass flow through the system. In order to be able to analyze these kind of systems with the help of state of the art computer codes the implemented numerical models for heat transfer, pressure drop or two phase flow structure must be able to predict the system performance in a wide parameter range. Goal of the program is to optimize the numerical models and therefore the performance of computer codes analyzing passive systems. Within the project the heat transfer capacity of a heat exchanger tube will be investigated. Therefore the tube will be equipped with detectors, both temperature and pressure, in several directions perpendicular to the tube axis to be able to resolve the angular heat transfer. In parallel the flow structure of a two phase flow inside and along the tube will be detected with the help of x-ray tomography. The water cooling outside of the tube will be realized by forced convection. It will be possible to combine the flow structure measurement with an angular resolved heat transfer for a wide parameter range. The test rig is set up at the TOPLFOW facility at the Helmholtz-Zentrum Dresden-Rossendorf (HZDR), so that it will be possible to vary the pressure between 5 and 70 bar. The steam mass content will be varied between 0 and 100 percent. The results will be compared to the large scaled Emergency Condenser Tests performed at the INKA test facility in Karlstein (Germany). The paper will explain the test setup and the status of the project will be presented. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Experimental characterization of pressure drops and channel instabilities in helical coil SG tubes

Helical tube heat exchangers provide better heat transfer characteristics, an improved capability to accommodate stresses due to thermal expansions and a more compact design with respect to straight tube heat exchangers. For these advantages they are considered as an option for the Steam Generator (SG) of many new reactor projects of Generation III+ and Generation IV. In particular, their compactness fits well with the requirements of Small-medium Modular Reactors (SMRs) of integral design, where all the primary system components are located inside the reactor vessel. In this framework, thermal hydraulics of helical pipes has been studied in recent years by Politecnico di Milano in different experimental campaigns. Experiments have been carried out in a full-scale open loop test facility installed at SIET labs in Piacenza (Italy)), to simulate the SG of a typical SMR. The facility includes two helical pipes (1 m coil diameter, 32 m length, 8 m height), connected via lower and upper headers. Following recently completed experimental campaigns dedicated to pressure drops and density wave instabilities, this paper deals with a new experimental campaign focused on both pressure drops (single-phase flow and two-phase flow, laminar and turbulent regimes) and flow instabilities. The availability of a large number of experimental data, in particular on two-phase flow, is of fundamental interest for correlation development, model validation and code assessment. Two-phase pressure drops have been measured in adiabatic conditions, ranging from 200 to 600 kg/m{sup 2}s for the mass flux, from 30 to 60 bar for the pressure and from 0.1 to 1.0 for the flow quality. The channel characteristics mass flow rate - pressure drop has been determined experimentally in the range 10-40 bar, varying the mass flow rate at a fixed value of the thermal flux. In addition, single-phase pressure drops have been measured in both laminar and turbulent conditions. Density wave instabilities have been studied at mass flux from 100 to 400 kg/m{sup 2}s and pressure from 10 to 20 bar, to confirm the particular behavior of the stability boundary in helical geometry at low pressure and low mass flow rate. Finally, starting from the unstable regions identified from the experimental channel characteristics, Ledinegg type instabilities have been investigated to drawn stability maps with complete stable and unstable regions in the dimensionless plane N sub-N pch. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

The technical and economic impact of minor actinide transmutation in a sodium fast reactor

Within the frame work of the French National Act of June 28, 2006 pertaining to the management of high activity, long-lived radioactive waste, one of the proposed processes consists in transmuting the Minor Actinides (MA) in the radial blankets of a Sodium Fast Reactor (SFR). With this option, we may assess the additional cost of the reactor by comparing two SFR designs, one with no Minor Actinides, and the other involving their transmutation. To perform this exercise, we define a reference design called SFRref, of 1500 MWe that is considered to be representative of the Reactor System. The SFRref mainly features a pool architecture with three pumps, six loops with one steam generator per loop. The reference core is the V2B core that was defined by the CEA a few years ago for the Reactor System. This architecture is designed to meet current safety requirements. In the case of transmutation, for this exercise we consider that the fertile blanket is replaced by two rows of assemblies having either 20% of Minor Actinides or 20% of Americium. The assessment work is performed in two phases. - The first consists in identifying and quantifying the technical differences between the two designs: the reference design without Minor Actinides and the design with Minor Actinides. The main differences are located in the reactor vessel, in the fuel handling system and in the intermediate storage area for spent fuel. An assessment of the availability is also performed so that the impact of the transmutation can be known. - The second consists in making an economic appraisal of the two designs. This work is performed using the CEA's SEMER code. The economic results are shown in relative values. For a transmutation of 20% of MA in the assemblies (S/As) and a hypothesis of 4 kW allowable for the washing device, there is a large external storage demanding a very long cooling time of the S/As. In this case, the economic impact may reach 5% on the capital part of the Levelized Unit Electricity Cost (LUEC). A diminished concentration at 10% of MA, reduces the size of the external storage and the cooling time of the assemblies becomes compatible with the management of the irradiated fuel. Even with a low allowable power for the washing device, the economic impact on the capital cost is less than 2.5%. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Ways of improvement for the materials of sodium cooled fast reactors

The French sodium cooled prototype reactor ASTRID will take into account 'Generation IV' requirements, especially a long operational life-time (60 years) and a high efficiency. The good behavior of austenitic steel AISI316L(N), should be confirmed for a use, in moderately irradiated and unirradiated parts of ASTRID. Parts recovered from dismantled French sodium-cooled reactors will be characterized. Further experiments must be carried out concerning ageing of these components. Other materials will be chosen for fuel wrapping and cladding, in order to reduce creep and swelling under irradiation, (either conventional, or oxide-dispersed strengthened steels (ODSS). Corrosion of ODSS in the presence of sodium needs a serious assessment The lifetime of primary pumps components made of Duplex steels should also be assessed. The disruptions in steam generator tubes should be minimized and controlled; therefore, optimised designs and geometries must be established before defining the corresponding materials. Either Modified 9Cr1Mo or Incoloy 800H, might be candidates;it will be necessary to check whether austenitic steels are compatible with Modified 9Cr1Mo or Incoloy 800H in the same circuit. For all materials, the best manufacturing processes must be combined with thermal, mechanical treatments; calculations of phase diagrams (CALPHAD) might be used to optimise both treatments and chemical compositions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Flow Accelerated Erosion-Corrosion (FAC) considerations for secondary side piping in the AP1000{sup R} nuclear power plant design

The issue of Flow Accelerated Erosion-Corrosion (FAC) in power plant piping is a known phenomenon that has resulted in material replacements and plant accidents in operating power plants. Therefore, it is important for FAC resistance to be considered in the design of new nuclear power plants. This paper describes the design considerations related to FAC that were used to develop a safe and robust AP1000{sup R} plant secondary side piping design. The primary FAC influencing factors include: - Fluid Temperature - Pipe Geometry/layout - Fluid Chemistry - Fluid Velocity - Pipe Material Composition - Moisture Content (in steam lines) Due to the unknowns related to the relative impact of the influencing factors and the complexities of the interactions between these factors, it is difficult to accurately predict the expected wear rate in a given piping segment in a new plant. This paper provides: - a description of FAC and the factors that influence the FAC degradation rate, - an assessment of the level of FAC resistance of AP1000{sup R} secondary side system piping, - an explanation of options to increase FAC resistance and associated benefits/cost, - discussion of development of a tool for predicting FAC degradation rate in new nuclear power plants. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Mercury Reduction and Removal from High Level Waste at the Defense Waste Processing Facility - 12511

The Defense Waste Processing Facility processes legacy nuclear waste generated at the Savannah River Site during production of enriched uranium and plutonium required by the Cold War. The nuclear waste is first treated via a complex sequence of controlled chemical reactions and then vitrified into a borosilicate glass form and poured into stainless steel canisters. Converting the nuclear waste into borosilicate glass is a safe, effective way to reduce the volume of the waste and stabilize the radionuclides. One of the constituents in the nuclear waste is mercury, which is present because it served as a catalyst in the dissolution of uranium-aluminum alloy fuel rods. At high temperatures mercury is corrosive to off-gas equipment, this poses a major challenge to the overall vitrification process in separating mercury from the waste stream prior to feeding the high temperature melter. Mercury is currently removed during the chemical process via formic acid reduction followed by steam stripping, which allows elemental mercury to be evaporated with the water vapor generated during boiling. The vapors are then condensed and sent to a hold tank where mercury coalesces and is recovered in the tank's sump via gravity settling. Next, mercury is transferred from the tank sump to a purification cell where it is washed with water and nitric acid and removed from the facility. Throughout the chemical processing cell, compounds of mercury exist in the sludge, condensate, and off-gas; all of which present unique challenges. Mercury removal from sludge waste being fed to the DWPF melter is required to avoid exhausting it to the environment or any negative impacts to the Melter Off-Gas system. The mercury concentration must be reduced to a level of 0.8 wt% or less before being introduced to the melter. Even though this is being successfully accomplished, the material balances accounting for incoming and collected mercury are not equal. In addition, mercury has not been effectively purified and collected in the Mercury Purification Cell (MPC) since 2008. A significant cleaning campaign aims to bring the MPC back up to facility housekeeping standards. Two significant investigations are being undertaken to restore mercury collection. The SMECT mercury pump has been removed from the tank and will be functionally tested. Also, research is being conducted by the Savannah River National Laboratory to determine the effects of antifoam addition on the behavior of mercury. These path forward items will help us better understand what is occurring in the mercury collection system and ultimately lead to an improved DWPF production rate and mercury recovery rate. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Passive acoustic leak detection for sodium cooled fast reactors using hidden Markov models

Acoustic leak detection for steam generators of sodium fast reactors have been an active research topic since the early 1970's and several methods have been tested over the years. Inspired by its success in the field of automatic speech recognition, we here apply hidden Markov models (HMM) in combination with Gaussian mixture models (GMM) to the problem. To achieve this, we propose a new feature calculation scheme, based on the temporal evolution of the power spectral density (PSD) of the signal. Using acoustic signals recorded during steam/water injection experiments done at the Indira Gandhi Centre for Atomic Research (IGCAR), the proposed method is tested. We perform parametric studies on the HMM+GMM model size and demonstrate that the proposed method a) performs well without a priori knowledge of injection noise, b) can incorporate several noise models and c) has an output distribution that simplifies false alarm rate control. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Education and training in the field of nuclear instrumentation and measurement: CEA/INSTN (National Institute for Nuclear Sciences and Technologies) strategy to improve and develop new pedagogical tools and methods

Part of the French Alternative Energies and Atomic Energy Commission (CEA), the National Institute for Nuclear Science and Technology (INSTN) is a higher education institution whose mission is to provide students and professionals a high level of scientific and technological qualification in all disciplines related to nuclear energy applications. In this frame, INSTN carries out education and training (E and T) programs in nuclear instrumentation and radioprotection. Its strategy has always been to complete theoretical courses by training courses and laboratory works carried out on an extensive range of training tools that includes a large panel of nuclear instrumentation as well as software applications. Since its creation in 1956, the INSTN has conducted both education and vocational programs on ionizing radiation detection. An extensive range of techniques have commonly been used during practical works with students and employees of companies who need to get the knowledge and specialization in this field. Today, the INSTN is mainly equipped with usual detectors and electronics in large numbers in order to be able to accommodate up to 48 trainees at the same time in two classrooms, with only two trainees for one workstation in order to optimize their learning. In the field of the neutron detection systems, the INSTN has strongly developed its offer taking advantage of the use of research reactors, such as ISIS reactor (700 kW) at Saclay. The implementation of neutron detection systems specific to the courses offers a unique way of observing and analysing the signal coming from neutron detectors, as well as learning how to set the parameters of the detection system in real conditions. Providing the trainees with an extensive overview of each part of the neutron monitoring instrumentation apply to a nuclear reactor, hands-on measurements on the ISIS reactor play a major role in ensuring a practical and comprehensive understanding of the neutron detection system and their integration in the safety system of nuclear reactors. It also gives a solid background for the follow up and the development of the neutron detection systems. Another field of activity of the INSTN is the development of new teaching tools using software applications. We describe hereafter two applications that have recently been developed by the institute, O.S.I.R.I.S and DOSIMEX. The INSTN and the company OREKA have developed an innovative teaching tool named O.S.I.R.I.S. The tool is built on a virtual 3D environment in which users operate in a totally free way. The action is located in a pressurised water reactor in the environment of a steam generator building. The users are students or professionals who want to learn and train on concrete situations on how to protect the workers against radiation. Through the use of this serious game, users must: establish a predictive dose evaluation, implement the principles of radiation protection, perform the dose result of the operation, analyse the gap between the predictive collective dose and the collective dose achieved, and decide about the lessons to provide feedback. Taking into account the increasing constraints related to radiation protection, it is essential to use computer codes to evaluate the dose rates generated by sources of ionizing radiation. Many powerful codes exist but are often relatively complex to implement and are dedicated to experts. In addition, these codes are often 'black boxes' that does not allow the user to understand the underlying physics. The objective of DOSIMEX codes developed by the INSTN is to give, by the means of a VBA code with a simple interface, the ability to easily conduct a wide range of radiological exposure situations: calculation of gamma, beta, neutron equivalent dose rate, calculation of internal contamination and atmospheric transfer.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Development of FeCrAl-ODS steels for ATF cladding

As an enhanced accident tolerant fuel (ATF) for LWR at severe accident condition, FeCrAl-ODS ferritic steels have been developed, and their applicability is being studied under the Japanese MEXT program. Alloying design of FeCrAl-ODS ferritic steels has been conducted, focusing on effect of mutual interaction of Cr-Al on oxidation resistance and α'-embrittlement. Effects of co-addition of Zr and excessive oxygen have also been studied to improve high-temperature strength. The ring tensile test of the manufactured claddings showed that tensile strength increases by Zr-addition and Zr-added FeCrAl-ODS steels have the highest strength among the existed steels at entire temperature range from room temperature up to 1000 Celsius degrees. Although Zr-addition sometimes leads to growth of alumina scale by rapid inward diffusion of oxygen through zirconia inside alumina, oxidation rate of Zr-added FeCrAl-ODS steels was successfully reduced to the level of Zr-free specimen by controlling oxygen content to form Y{sub 2}Zr{sub 2}O{sub 7} inside ODS steel matrix. This finding makes possible to improve not only high-temperature strength but also steam oxidation resistance in the Zr-added FeCrAl-ODS steels. (authors)

36 MATERIALS SCIENCE↗

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