SEARCH · Engineering Papers

Results for “REACTOR INSTRUMENTATION”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

SourceResearch connectionFrom yearThrough year

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 721 records · Page 40

All sources NASA NTRS DOE OSTI

DOE OSTI2023

Volume 9: Critical Facility with Add-On Ion Beam

Completion of the originally planned critical facility has been considered as part of the High Flux Isotope Reactor (HFIR) Sustaining and Enhancing Neutron Science Initiative at Oak Ridge National Laboratory (ORNL). A working group of ORNL staff members was formed to develop the idea and to recommend one or more configurations to best support future HFIR operations and scientific capacities. HFIR was designed with a critical pool in the reactor bay that was projected to be outfitted as a critical facility. The primary purpose of the planned critical facility was to measure the subcritical worth of fresh fuel elements to support startup requirements. However, the Y-12 National Security Complex already housed a critical facility that was used for this purpose. This report presents the working group’s efforts, including determination of the proposed critical facility’s high-impact benefits. A generic low-power critical facility would be employed for reactor physics measurements, code and data validation, reactor operator and staff training, and education. This facility would be instrumental in supporting current HFIR operations, conversion of HFIR to low-enriched uranium, existing light water reactor operations, and advanced reactor development and deployment.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

Post-irradiation Examinations of High Burnup PWR Rods

The High Burnup Spent Fuel Data Project (herein referred to as the “project”), sponsored by the US Department of Energy (DOE) Office of Nuclear Energy (NE), is focused on experimentally evaluating the effects of long-term storage and transportation on high burnup (HBU) (defined as >45 gigawatt days per metric ton uranium) pressurized water reactor (PWR) spent nuclear fuel (SNF) [1]. To support this project, in the fall of 2017, an instrumented bolted-lid cask (project cask) at the North Anna reactor site was loaded with intact, HBU PWR fuel assemblies (project assemblies) having four different kinds of cladding and will be reopened in approximately 10-years for inspection. To support pre- and post-storage SNF rod characterization, 25 “sister rods” that are similar to the rods being stored in the project cask were extracted from seven different fuel assemblies at the North Anna nuclear power plant for post-irradiation examination (PIE).The sister rod testing program has multiple objectives, including: 1) Baseline testing to provide essential information on the physical condition and properties of the cladding and the fuel in the cladding prior to the loading, drying, and long-term dry storage process (analogous to an experiment control group). Similar tests are expected to be performed at the end of the 10-year storage period on rods extracted from the cask to identify any changes in the properties of the fuel rods that may have occurred during the dry storage period; and 2) Tests to provide material property and physical data using accelerated separate effects tests (SETs) and small-scale tests (SSTs) to explore a wide range of conditions that may occur for other dry storage systems and/or while the fuel is being transported after aging.Testing will be focused on understanding overall SNF rod strength and durability, and be comprised of both composite fuel and defueled clad testing with a focus on measuring cladding material and mechanical properties. Nondestructive testing is currently underway and plans are being updated to support the initial set of destructive tests that are expected to begin by early 2018. While the data generated can be used for multiple purposes, a primary goal of the sister rod PIE data and associated measured mechanical properties are expected to support existing fuel storage licensing and relicensing activities by addressing identified knowledge gaps, and enhance the technical bases for post-storage transportation, handling, and consolidation activities [1]. This paper presents results obtained from the sister rod testing to date, and describes some of the planned measurements, equipment, and procedures that will be implemented for the initial set of destructive examinations.

Montgomery, Rose↗

DOE OSTI2018

Post-Irradiation Examinations of High Burnup PWR Fuel Rods - Initial Results

Bevard, Bruce Balkcom↗

DOE OSTI2018

Post-irradiation Examinations of High Burnup PWR Rods - 18456

The High Burnup Spent Fuel Data Project (herein referred to as the 'project'), sponsored by the US Department of Energy (DOE) Office of Nuclear Energy (NE), is focused on experimentally evaluating the effects of long-term storage and transportation on high burnup (HBU) (defined as >45 gigawatt days per metric ton uranium) pressurized water reactor (PWR) spent nuclear fuel (SNF) [1]. To support this project, in the fall of 2017, an instrumented bolted-lid cask (project cask) at the North Anna reactor site was loaded with intact, HBU PWR fuel assemblies (project assemblies) having four different kinds of cladding and will be reopened in approximately 10-years for inspection. To support pre- and post-storage SNF rod characterization, 25 'sister rods' that are similar to the rods being stored in the project cask were extracted from seven different fuel assemblies at the North Anna nuclear power plant for post-irradiation examination (PIE). The sister rod testing program has multiple objectives, including: 1) Baseline testing to provide essential information on the physical condition and properties of the cladding and the fuel in the cladding prior to the loading, drying, and long-term dry storage process (analogous to an experiment control group). Similar tests are expected to be performed at the end of the 10-year storage period on rods extracted from the cask to identify any changes in the properties of the fuel rods that may have occurred during the dry storage period; and 2) Tests to provide material property and physical data using accelerated separate effects tests (SETs) and small-scale tests (SSTs) to explore a wide range of conditions that may occur for other dry storage systems and/or while the fuel is being transported after aging. Testing will be focused on understanding overall SNF rod strength and durability, and be comprised of both composite fuel and defueled clad testing with a focus on measuring cladding material and mechanical properties. Nondestructive testing is currently underway and plans are being updated to support the initial set of destructive tests that are expected to begin by early 2018. While the data generated can be used for multiple purposes, a primary goal of the sister rod PIE data and associated measured mechanical properties are expected to support existing fuel storage licensing and relicensing activities by addressing identified knowledge gaps, and enhance the technical bases for post-storage transportation, handling, and consolidation activities [1]. This paper presents results obtained from the sister rod testing to date, and describes some of the planned measurements, equipment, and procedures that will be implemented for the initial set of destructive examinations. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

I&C System Sensors for Advanced Nuclear Reactors

As advanced reactor designers proceed with further research, design implementation, testing, and deployment, it is now time to address their instrumentation and control (I&C) needs. First, we must identify the process variables that must be measured to control an advanced reactor and monitor its safety. As with any nuclear power plant, it is safe to assume that temperature, pressure, flow, level, and neutron flux are the predominant variables to be measured. As a second step, the range of process values must be established as well as the level of accuracy, redundancy, and speed of response of sensors that must be deployed to measure these values. The harsh environments of advanced reactors such as high temperatures, high levels of nuclear radiation, potential for corrosion, and limited access to sensors for maintenance are just a few examples of instrumentation challenges. This is compounded by the relatively small market for new sensors that must be designed, developed, and qualified for advanced reactors. That is, the cost of these activities can deter major sensor manufacturers from entering the advanced reactor market. Furthermore, the operating cycles of most advanced reactors are expected to be much longer than conventional plants adding to the need for instrumentation that can maintain calibration and response time over extended intervals. The fact that there are now a variety of advanced reactors under design and development adds to the complexity of instrumentation that may differ for various types of advanced plants. A few examples of the most commonly discussed advanced reactors are reviewed in this paper followed by a brief discussion of their instrumentation needs.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2020

Evaluation of Low Dose Silicon Carbide Temperature Monitors

Thermocouples are generally used to provide real-time temperature indications in instrumented tests performed at materials and test reactors. Melt wires or paint spots are often included in such tests as an independent technique of detecting peak temperatures incurred during irradiation. In addition, less expensive static capsule tests, which have no leads attached for real-time data transmission, often rely on melt wires and paint spots as a post-irradiation technique for peak temperature indication. Unfortunately, these techniques are limited in that they can only detect whether a single temperature is or is not exceeded. Silicon carbide (SiC) monitors are advantageous because a single monitor can be used to determine the peak temperature reached within a relatively broad range (200 – 800°C). Although the use of SiC monitors was proposed more than five decades ago, the ultimate performance limits of this technique are not fully understood. The Nuclear Science User Facilities (NSUF) is the United States Department of Energy Office of Nuclear Energy's only designated nuclear energy user facility. Its mission is to provide nuclear energy researchers access to world-class capabilities and to facilitate the advancement of nuclear science and technology. This mission is supported by providing access to state-of-the-art experimental irradiation testing, post irradiation examination facilities, and high performance computing capabilities as well as technical and scientific assistance for the design and execution of projects. As part of an NSUF project, low dose silicon carbide monitors were irradiated in the Belgian Reactor 2 and were then evaluated both at the SCK•CEN and at Idaho National Laboratory's High Temperature Test Laboratory to determine their peak temperature achieved during irradiation. The technical significance of this work was that the monitors were irradiated to a dose that was significantly less than recommended in published literature. This paper will discuss the evaluation process, the irradiation test, and the performance of the low dose silicon carbide temperature monitors.

36 MATERIALS SCIENCE↗

DOE OSTI2017

Frequency spectrum analysis code for nuclear noise based on improved wavelet algorithm - 337

The core of the narrow flow channel generally faced the problem of local bubble boiling with the temperature rise of the nuclear fuel element, it is necessary and meaningful to answer and deal with the problem on the technical level. Reactor noise analysis begins with the study of the physical characteristics of the core, after nearly 30 years of research and development it has become a powerful tool in the auxiliary monitoring of the power reactor operating. By using this noise frequency spectrum analysis technology, the reactor dynamic parameter such as prompt neutron decay constant can be measured. Also, through such system, the frequency characteristics of reactor can be real-time monitored, which is helpful for safety operation and fault diagnosis. Spectrum analysis technology in the Fourier analysis era is now barely available, the reason is that the local spectral characteristics of this technology are not obvious, but this defect can be greatly improved by wavelet analysis technique, however, the traditional wavelet analysis will introduce the problem of spectral confusion. To solve the problem, based on the improved wavelet algorithm, frequency spectrum analysis code for nuclear noise is developed,from the perspective of fine spectral analysis, the local boiling accident of the coolant can be prejudged and avoid. Based on the analysis and test of experimental data,The results show that the calculation ability of the code is in accordance with the expected, it can be applied in the operation of research reactor. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

On-line monitoring for static and dynamic performance verification of I and C systems at Sizewell B nuclear power station - 329

Plant operations and maintenance (O and M) activities account for nearly 75% of the generation cost of a modern nuclear power plant (NPP). This is largely due to time-based maintenance strategies that are expensive, time-consuming, prone to error, and sometimes detrimental to plant equipment. In contrast, condition-based maintenance via on-line monitoring (OLM) can reduce costs, increase equipment availability, and improve safety. OLM uses signals from existing plant sensors to verify the performance of plant instrumentation and equipment while the plant is operating. The Sizewell B nuclear power station in the United Kingdom, a four-loop pressurized water reactor (PWR), uses OLM techniques for several condition-based performance verification activities including pressure, level, and flow transmitter calibration monitoring, resistance temperature detector (RTD) cross calibration, and dynamic performance monitoring of pressure, level, and flow sensing systems. The implementation of OLM has enabled the plant to extend the calibration intervals of its pressure, level, and flow transmitters from 18 to 54 months and has significantly reduced the amount of critical path time needed to perform RTD cross calibration testing (up to 36 hours). Dynamic performance monitoring of its pressure, level, and flow transmitters with OLM has enabled the plant to monitor transmitter and sensing line response degradation while the plant is operating at 100% power and has dramatically reduced testing time versus the traditional hydraulic ramp testing method. The Sizewell B OLM implementation provides valuable insights into on how to overcome the technical and regulatory challenges for widespread OLM adoption by the nuclear industry. This paper describes the technical and regulatory aspects of this implementation and includes detailed examples of static and dynamic performance testing results to demonstrate the types of capabilities of OLM. Analyzing existing signals with OLM technologies can provide significant cost and safety benefits for the existing fleet of nuclear power plants. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2012

Linear variable differential transformer (LVDT)-based elongation measurements in Advanced Test Reactor high temperature irradiation testing

New materials are being considered for fuel, cladding and structures in next generation and existing nuclear reactors. These materials can undergo significant dimensional and physical changes during high temperature irradiations. Currently, such changes are determined by repeatedly irradiating a specimen for a specified period of time in the Advanced Test Reactor (ATR) and then removing it from the reactor for evaluation. The labor and time to remove, examine and return irradiated samples for each measurement make this approach very expensive. In addition, such techniques provide limited data and may disturb the phenomena of interest. To resolve these issues, an instrumented creep testing capability is being developed for specimens irradiated under pressurized water reactor coolant conditions in the ATR at the Idaho National Laboratory (INL). This paper reports the status of INL efforts to develop this testing capability. In addition to providing an overview of in-pile creep test capabilities available at other test reactors, this paper focuses on efforts to design and evaluate a prototype test rig in an autoclave at INL’s High Temperature Test Laboratory.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2015

Training courses on neutron detection systems on the ISIS research reactor: on-site and through internet training

Today, ISIS research reactor is an essential tool for Education and Training programs organized by the National Institute for Nuclear Science and Technology (INSTN) from CEA. In the field of nuclear instrumentation, the INSTN offers both, theoretical courses and training courses on the use of neutron detection systems taking advantage of the ISIS research reactor for the supply of a wide range of neutron fluxes. This paper describes the content of the training carried out on the use of neutron detectors and detection systems, on-site or remote. The ISIS reactor is a 700 kW open core pool type reactor. The facility is very flexible since neutron detectors can be inserted into the core or its vicinity, and be used at different levels of power according to the needs of the course. Neutron fluxes, typically ranging from 1 to 10{sup 12} n/cm{sup 2}.s, can be obtained for the characterisation of the neutron detectors and detection systems. For the monitoring of the neutron density at low level of power, the Instrumentation and Control (I and C) system of the reactor is equipped with two detection systems, named BN1 and BN2. Each way contains a fission chamber, type CFUL01, connected to an electronic system type SIREX.The system works in pulse mode and exhibits two outputs: the counting rate and the doubling time. For the high level of power, the I and C is equipped with two detection systems HN1 and HN2.Each way contain a boron ionization chamber (type CC52) connected to an electronics system type SIREX. The system works in current mode and has two outputs: the current and the doubling time. For each mode, the trainees can observe and measure the signal at the different stages of the electronic system, with an oscilloscope. They can understand the role of each component of the detection system: detector, cable and each electronic block. The limitation of the detection modes and their operating range can be established from the measured signal. The trainees can also modify the settings of the electronic system, such as the high voltage and the discrimination level in order to obtain all the characteristic curves of the detectors. These curves are used to define the right setting of the electronic system and to discuss the expected degradation of the detector signal resulting from the detector damage under the integrated neutron and gamma fluxes. Moreover, in addition to the study of the neutron detection systems itself, the integration of the measurements made by these detection systems in the logic of the safety system of the nuclear reactor is also addressed. 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. In addition to on-reactor training, Internet Reactor Laboratory capability has been implemented on the ISIS reactor in 2014. For the Internet Reactor Laboratory an extensive video conference system has been implemented on ISIS reactor. The system includes 4 cameras and the transmission of the video signal given by the supervision system of the reactor which records and processes the data of the reactor. According to the pedagogic needs during the training courses, the lecturer on the ISIS reactor chooses to broadcast the relevant information at each stage of the course. For example, graph showing the histogram of the counting and current as a function of the time, or the electrical signal observed on the oscilloscope, can be broadcasted trough internet. By interacting through the video conference, the remote classroom is able to ask for changes in the reactor power or settings of the detection systems. They can also ask for the broadcast of some particular information. At the guest institution, the information is displayed in two parts or screens, as shown in the Figure 3. Concerning the interaction with - and the feedback from - the remote classroom, the camera of the video system in the remote classroom is used to ensure the contact between the trainees and the lecturer and reactor operators. Thus, the Internet Reactor Laboratory is complementary to the on reactor training courses. It allows distant learning, reducing the overall cost of the course when this is necessary. It can efficiently be used for the development of the human resources needed by the nuclear industry and the nuclear programs in countries without research reactors.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Instrumentation and Data Analysis Supporting the Transient Rod Pneumatic System Design Study at the Annular Core Research Reactor.

Abstract not provided.

Arnold, James F.↗

DOE OSTI2014

Irradiation Testing of Ultrasonic Transducers

Ultrasonic technologies offer the potential for high accuracy and resolution in-pile measurement of a range of parameters, including geometry changes, temperature, crack initiation and growth, gas pressure and composition, and microstructural changes. Many Department of Energy-Office of Nuclear Energy (DOE-NE) programs are exploring the use of ultrasonic technologies to provide enhanced sensors for in-pile instrumentation during irradiation testing. For example, the ability of single, small diameter ultrasonic thermometers (UTs) to provide a temperature profile in candidate metallic and oxide fuel would provide much needed data for validating new fuel performance models. Other efforts include an ultrasonic technique to detect morphology changes (such as crack initiation and growth) and acoustic techniques to evaluate fission gas composition and pressure. These efforts are limited by the lack of existing knowledge of ultrasonic transducer material survivability under irradiation conditions. For this reason, the Pennsylvania State University (PSU) was awarded an Advanced Test Reactor National Scientific User Facility (ATR NSUF) project to evaluate promising magnetostrictive and piezoelectric transducer performance in the Massachusetts Institute of Technology Research Reactor (MITR) up to a fast fluence of at least 1021 n/cm2 (E> 0.1 MeV). The goal of this research is to characterize magnetostrictive and piezoelectric transducer survivability during irradiation, enabling the development of novel radiation tolerant ultrasonic sensors for use in Material and Test Reactors (MTRs). As such, this test will be an instrumented lead test and real-time transducer performance data will be collected along with temperature and neutron and gamma flux data. The current work bridges the gap between proven out-of-pile ultrasonic techniques and in-pile deployment of ultrasonic sensors by acquiring the data necessary to demonstrate the performance of ultrasonic transducers.

harsh-environment sensors↗

DOE OSTI2017

Data analysis for valve leak detection of nuclear power plant safety critical components - 336

A nuclear plant consists of several safety-related components. These safety-related components are tested periodically in order to verify their functionality when they are called upon to perform. The data acquired during periodic testing provides information about possible degradation of the equipment or device under test. This research is focused on leak monitoring techniques to provide diagnostic information on the determination of valve aging and degradation. Several commercially available valve diagnostic-monitoring methods are evaluated, especially for the techniques based on measurements of acoustic emission, and accelerometer. The check valve is one of typical components being extensively used in safety systems of nuclear power plants. The failures of check valves have resulted in significant maintenance efforts, on occasion, have resulted in water hammer, over-pressurization of low-pressure systems, and damage to flow system components. In this paper, acoustic emission signals due to leak from check valves with artificial leakage and worn parts are investigated both analytically and experimentally. The results of advanced signal processing and de-noising techniques also have been identified. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2011

Neutrons for Catalysis: A Workshop on Neutron Scattering Techniques for Studies in Catalysis

This report summarizes the Workshop on Neutron Scattering Techniques for Studies in Catalysis, held at the Spallation Neutron Source (SNS) at Oak Ridge National Laboratory (ORNL) on September 16 and 17, 2010. The goal of the Workshop was to bring experts in heterogeneous catalysis and biocatalysis together with neutron scattering experimenters to identify ways to attack new problems, especially Grand Challenge problems in catalysis, using neutron scattering. The Workshop locale was motivated by the neutron capabilities at ORNL, including the High Flux Isotope Reactor (HFIR) and the new and developing instrumentation at the SNS. Approximately 90 researchers met for 1 1/2 days with oral presentations and breakout sessions. Oral presentations were divided into five topical sessions aimed at a discussion of Grand Challenge problems in catalysis, dynamics studies, structure characterization, biocatalysis, and computational methods. Eleven internationally known invited experts spoke in these sessions. The Workshop was intended both to educate catalyst experts about the methods and possibilities of neutron methods and to educate the neutron community about the methods and scientific challenges in catalysis. Above all, it was intended to inspire new research ideas among the attendees. All attendees were asked to participate in one or more of three breakout sessions to share ideas and propose new experiments that could be performed using the ORNL neutron facilities. The Workshop was expected to lead to proposals for beam time at either the HFIR or the SNS; therefore, it was expected that each breakout session would identify a few experiments or proof-of-principle experiments and a leader who would pursue a proposal after the Workshop. Also, a refereed review article will be submitted to a prominent journal to present research and ideas illustrating the benefits and possibilities of neutron methods for catalysis research.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Fuel Motion Monitoring System at TREAT – FY2017 Instrumentation Summary

This document is a report summarizing FY2017 efforts related to the Transient Reactor Test Facility (TREAT) Fuel Motion Monitoring System, "hodoscope," refurbishment campaign. TREAT is an air-cooled, graphite-moderated, heterogeneous test facility designed to evaluate reactor fuels and structural materials under conditions that simulate various types of transient overpower and under-cooling situations in a nuclear reactor. Fuel meltdowns, metal-water reactions, thermal interaction between overheated fuel and coolant, and the transient behavior of ceramic fuel for high-temperature systems can be studied. A key instrument that monitors fuel motion as these events take place is the fast neutron hodoscope. The hodoscope is designed to allow for pre-, during-, and post- transient imaging of fuel within a test loop in the center of the reactor core. The preferred technique for imaging is detection of fast neutrons produced by fission in the test fuel, travelling unmoderated through an empty row of core fuel assemblies and then through a multi-slot collimator to a 360-detector array. Due to the age of the instrument, a refurbishment of the hodoscope’s fast neutron detection capability is necessary to support real-time, fuel-position experimental data collection. One particular challenge in this regard is related to the refurbishment of the fast-neutron-sensing detector elements in the detector array. This report summarizes major activities and accomplishments related to the FMMS refurbishment in FY2017. Of particular note was work related to identifying options for new proton recoil scintillator wafer and initial examination of the TREAT proton recoil proportional counter detectors.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2011

Ultrasound in lead-bismuth eutectic

The Belgian Nuclear Research Centre (SCK.CEN) is in the process of designing MYRRHA, a new multi-purpose irradiation facility to replace the ageing BR2. MYRRHA is a fast spectrum reactor cooled with lead-bismuth eutectic (LBE). As liquid metal is opaque to visual light, ultrasonic measurement techniques are selected to fulfill essential tasks that, according to our assessment, will be demanded by licensing authorities, in particular: fuel assembly identification and localization of a lost fuel assembly. To that end, a considerable research effort at SCK.CEN is devoted to study ultrasonic propagation in LBE. As ultrasonic experiments in LBE are elaborate and expensive to set up, we are particularly interested in to what extent experiments in water can be extrapolated to LBE - one of the main focuses of this article. We describe and present results of a first experiment with this goal which shows that the signal to noise ratio is better in LBE and that we even see small diffuse reflections up to 40 deg. off normal. On the other hand, we do not see internal reflections in stainless steel objects in LBE which we do in water. Therefore, we conclude that experiments in water can be used to validate algorithms for LBE on the condition that they do not rely on internal reflections. We also present solutions to tackle the essential tasks: fuel assembly identification and lost object localization. The requirements for the ultrasonic equipment implementing these solutions are also discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

Design and implementation of safety channels for the Idaho State University AGN-201m research reactor control console - 216

The purpose of the work described in this paper is to design, implement, and install new safety channels for the control console of the Idaho State University (ISU) AGN-201m Research Reactor. These channels will replace the current vacuum-tube driven system, which has been in operation for over 50 years at the ISU nuclear engineering laboratory. Due to the aging issues associated with this technology an entirely new control console is needed. There are 3 safety channels which serve as the monitors and safety systems for adequately shutting down the reactor should any limits be broken. These 3 safety channels work harmoniously with other systems within the new control console to ensure appropriate operation of the reactor. The safety channel replacements are composed of three complementary signal processing systems whose circuits monitor the operating parameters of the reactor. The key reactor information is shown on several independent meters corresponding to each of the three safety channels. Finally, the reactor trip points are set to correspond with the standards promulgated by the Nuclear Regulatory Commission (NRC) to ensure safe operation of the ISU AGN-201m Research Reactor. The interconnections between each of these systems ensure that if any Channel fails or otherwise breaks a limit the entire system will initiate a scram and the reactor will become adequately sub-critical. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2015

2013 Caliban and Prospero experiments for the study of the behavior of radiation protection instrumentations in case of a criticality accident - 14346

Experiments using the CALIBAN and PROSPERO reactors were performed in 2013 in the framework of a collaboration between IRSN and CEA. The objective was to provide data about the possibility to use information given by radiation protection instrumentations (RPI) in the case of a criticality accident. In that purpose, the behavior of RPI was studied for various kinetics of the reactors which simulated different starts of a criticality accident. The article presents the safety issues associated with these experiments, i.e. the possible use of information given by the RPI for the emergency response of a criticality accident (detection of the accident, triage of the personnel, follow-up of the accident and reentry). Then, one paragraph is dedicated to the presentation of the tested material. In addition, the CALIBAN and PROSPERO reactors are briefly described, followed by the performed experiments. The RPI are neither designed nor dedicated to be used in case of a criticality accident. However, the preliminary conclusion of the experiments performed in 2013 is that the tested RPI have a correct behavior for slow kinetic experiments performed with the PROSPERO reactor, but might give less reliable and accurate information for pulse experiments performed with the CALIBAN reactor. In the last case, they can nevertheless provide information that can be helpful for the emergency response and the limitation of the consequences of the criticality accident. However, if the RPI are included in the emergency response of a criticality accident (detection, triage, follow-up, reentry), emergency procedures should be adapted. Other experiments might be performed in order to confirm results and to extend the number of tested materials, in particular portable survey meters or contamination area monitors. (authors)

42 ENGINEERING↗