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At least 433 records · Page 24
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Transport and Treatment of Chapelcross Gas Ducts for Clearance and Recycling - 19662

In the summer of 2018, Cyclife, the EDF group's specialist radioactive waste management and decommissioning company, transported sixteen gas ducts and associated equipment from the shutdown Chapelcross nuclear power plant to treatment facilities in the UK and Sweden. The client requested a full removal, treatment and disposal service for approximately 820 te of low level metallic waste comprising Upper Gas Duct Assemblies (including bellows), Upper Gas Duct elbows, Filter Pots and Convoluted pipework. The complex scope involved heavy lifting, transport by road and sea, processing and recycling on a large scale, thereby necessitating involvement of a wide range of suppliers at different tiers all managed by Cyclife. The purpose of the entire project was to safely handle and treat all of the components for clearance and recycling into the public domain for the manufacture of new steel products. Most of the treatment took place in Cyclife's nuclear licensed metal treatment facility in Sweden. Cyclife has a long and unique experience of arranging safe transport of large contaminated components on road and sea. The metal treatment facility in Sweden can, without special arrangements, receive components up to 30 meters long with a weight of up to 400 tonnes for decontamination, size reduction and melting. The actual treatment of the components consisted of the following steps: a) Separation of a gas duct from the gantry, b) size reduction, radiological verification measurements and clearance of the gantry, c) stepwise segmentation of the gas duct and elbow including removal of asbestos, d) decontamination, e) melting, f) radiological assessment and clearance of the metal ingots, g) management of the secondary waste for return to UK. The result of the project, upon completed treatment of all the material, demonstrates a high recycling rate of the material. 5-10% of the original mass of material will have to disposed as radioactive waste. This number include all residues, untreatable material (asbestos, mineral wool, filters etc.) and two melt batches with a too high activity content for clearance. The paper takes the reader through the entire project from the initial planning to the clearance of the metal and the return of the secondary waste to UK for disposal. It will also present and discuss pros and cons for alternative management approaches in several perspectives. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

New Designed Containers for Activated ILW from NPP Decommissioning - 20208

The paper focuses on packaging of dismantled and fragmented reactor components into containers and their transport to their storage and/or disposal facility within the immediate dismantling strategy applied on decommissioning of nuclear power plant (NPP) finally shutdown after standard operation. The main purpose of the paper is to present the process of design, manufacture, testing and use of new transport and storage containers for activated intermediate level waste (ILW) from V1 NPP decommissioning at Jaslovske Bohunice site. Legislation conditions, container selection process, requirements on their design and manufacture, scope of associated licensing documentation and technical procedures applied up to their use on site are described in detail. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Optimization of the Post-Operational Phase on Two Belgian Multi-Unit Nuclear Power Plants: the Case of the Non-Fissile Irradiated Core Items - 20156

The current legal framework in Belgium foresees the progressive phase out of nuclear power between October 2022 (Doel 3) and December 2025 (Doel 2). Upon its definitive shutdown, each unit of the Tihange and Doel sites will enter a Post-Operational Phase (POP) and be prepared for its Decontamination and Decommissioning (D and D). Prior to obtaining the D and D license, the Operator Electrabel is legally required to remove any non-fissile irradiated core items stored in the deactivation pools. The non-fissile irradiated core items consist essentially of control rods, poison rods and source thimbles as well as thimble plugs and foreign materials irradiated during operation: - Their significant content in highly radiant radionuclides (up to 6 TBq of Co-60 per kg of irradiated material) renders all existing operational waste management processes inadequate due to insufficient biological shielding; - Their high concentrations in long-lived radionuclides call for their disposal in a geological repository for which no final design nor waste acceptance criteria are expected prior to 2050. Uncertainties in the Belgian energy supply and security, however, require the Operator to be prepared for a partial nuclear phase out, where one or more units would benefit from lifetime extension while the remaining units would undergo decommissioning. The present paper aims at presenting how Electrabel, in partnership with Tractebel, addressed this challenge by maximizing the use of synergies within the respective sites as well as between both sites themselves, all the while accounting for site specificities. The most recent results and state of progress of the project will be detailed and the first lessons learned will be shared. The project has been split in multiple tasks and phased as follows: - An inventory phase aimed at mapping the contents, origin, composition and history of the non-fissile irradiated core items; - A pre-characterization phase based on neutron activation models; - A waste sorting phase aimed at separating waste forms for which an evacuation route exists from those for which such route does not exist; - A feasibility phase aimed at exploring all possible scenarios for the management of non-fissile irradiated core items and identifying the optimal feasible solution for each site; - A preparation phase (currently ongoing), developing further the optimal solution and ensuring that back-up solutions are available for any foreseeable change of context (licensing issue, modification in the nuclear phase-out program, etc.) and initiating early contacts with potential subcontractors for segmentation works and cask manufacturers, as well as the Belgian regulatory body and waste management agency. This phase also foresees the investigation of destructive and non-destructive radiological measurements to support the detailed characterization of the waste forms; - A realization phase (future work). (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Feedback from International Decommissioning Planning Projects - 20269

Decommissioning of Nuclear Power Plants is an increasing reality in many countries with nuclear power. Already many years before final shutdown, planning of the decommissioning work is usually started in order to understand for example the costs, waste volume capacity and licensing work needed. In many countries it is required to have a final decommissioning plan before any decommissioning work is allowed to start. Westinghouse has been involved in international decommissioning planning for more than 30 years, where the first 20 years was mostly funding and preliminary plans due to that decommissioning was many years away. However, as many more plants are now moving into decommissioning, the focus has shifted to final decommissioning plans with budgets instead of funding plans, strategic decisions instead of choices, waste volume specifications instead of estimates and a defined approach instead of what ifs. Westinghouse has developed a methodology based on the planning and execution of many decommissioning projects in Europe, America and Asia. That international experience and understanding helps the customers to adopt a systematic planning process where strategy development is done with the end state in mind, where the goals are clearly outlined and the starting point described. The process and steps to reach the desired end state need to be clearly defined and all options/alternatives evaluated from a cost/benefit point of view. This is followed by a thorough and quantitative risk analysis in order to give solid recommendations and credible conclusions. One important feedback has been that real decommissioning experience from reference projects is invaluable for an efficient and timely progress in order not to spend unnecessary time and effort on understanding the barriers and links between different activities in decommissioning. Skilled project managers or engineers can drive the project forward, but without decommissioning experience previously gained, valuable lessons learned are not used and the same mistakes can be made again. Many utilities have invested in knowledge transfer from other decommissioning projects through site visits, exchange programs and supplier meetings, which is a valuable asset and way of implementing lessons learned from performed decommissioning projects into the planning process. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Decommissioning of the V1 Nuclear Power Plant Jaslovske Bohunice - 20270

The paper focuses on introducing the approach and method applied for decommissioning of the V1 nuclear power plant (further the 'V1 NPP') located in Jaslovske Bohunice in the Slovak Republic, to the conference participants. The Paper outlines activities and steps necessary to comply with the requirements contained in the decision of the Slovak government from 1999 regarding the early shutdown of the V1 NPP. The company responsible for the V1 NPP decommissioning is Jadrova a vyradovacia spolocnost, a.s., (further JAVYS) with the State as a sole shareholder. On the basis of the V1 NPP Decommissioning Conceptual Plan, out of four assessed options, JAVYS selected the immediate decommissioning option (the 'IDO') for the V1 NPP. On the basis of multi-criteria analysis, the IDO specified the approach of immediate and continuous dismantling of equipment followed by demolition of buildings and preparation of the site for its further potential use. The IDO also included the summary of technical, environmental, legislative information as well as the determination of assumed costs for decommissioning of the V1 NPP. Prior to commencement of the V1 NPP decommissioning, within the process of termination of operation, activities connected with de-fuelling, monitoring, processing of historical waste and, not to forget, the process of obtaining the license for the V1 NPP decommissioning were executed. JAVYS decided to divide the decommissioning process into two stages. In 2011, JAVYS obtained a license for the first stage of the V1 NPP decommissioning. Within this stage, dismantling of inactive equipment, demolitions of inactive buildings, monitoring of systems and preparatory works for the upcoming stage of decommissioning were implemented. In 2015, in line with the license for the 2. stage of the V1 NPP decommissioning, the activities of decontamination and dismantling of contaminated and active equipment as well as demolitions of remaining buildings began. In this stage radioactive waste (RAW) produced during the decommissioning is to be processed. After decontamination, the material fulfilling the free release criteria is to be released into the environment, the site is to be cleared from the scope of the Atomic Act and released from the surveillance of the national regulators. Future application of certain methods and techniques for decontamination and dismantling in similar environments might bring an added value to the operators. Paper includes a description of best practices for Nuclear Power Plant VVER type reactors decommissioning. Schedule comparison is given to show how the critical path of certain projects is being managed. Application of the International Standard for Decommissioning Costing (ISDC) methodology provides a uniform method which enables comparison of cost estimates to other nuclear facilities under decommissioning. The cost estimate for the V1 NPP decommissioning has been continuously updated based on more advanced and detailed information on the decommissioning activities and projects. Monitoring of progress of the V1 NPP decommissioning is performed via four key-performance indicators. The Earned Value Management (EVM) methodology is used for evaluating time and cost aspect of decommissioning activities while waste conditioning index is calculated by comparing the actually produced waste to planned values of waste for a respective monitored period. Radiological safety index is assessed by comparison of the personnel records on the individual yearly doses received by each respective employee to the boundary value. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

The coupled neutronics and thermo-fluid dynamics design characteristics of the XE-100 200 MW{sub TH} reactor - 18552

The Xe-100 is a high temperature helium-cooled, graphite moderated, pebble bed reactor with a 10 times through, multi-pass (MEDUL=MEhrfacDUrchLauf) fueling scheme. The design, characterized as a Generation IV reactor, is based on proven technology that has been demonstrated a number of times both experimentally and as commercial reactors. This proven foundation ensures high levels of design and technical readiness. In the overview provided of the design it is shown how the excess reactivity is limited by continuous refueling, whilst the selected core power density will ensure passive heat removal. In this way the intrinsic safety design features of a typical GEN IV design are guaranteed. It is shown that no upset event will lead to a significant amount of radiological release. The coupled core neutronic and thermo-fluid dynamics design for the Xe-100 is performed with the VSOP-A system of codes. For the equilibrium core, VSOP results demonstrate that the spherical fuel powers (maximum 1.9 kW) and operational temperatures (<900 deg. C) fall well within the envelope of the design criteria. Adequate reactivity control and long-term, cold shutdown are provided by two separate, diversely-actuated systems, while the overall negative reactivity temperature coefficient is illustrated over the total operational range. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Identification and repair of intermittent cable faults in nuclear power plants - 131

The nuclear industry uses cable circuits as an integral part of a plant's reactor protection system (RPS), instrumentation and control (I and C) and power distribution systems. The reliability of these systems is critical for continued operation of the world's aging nuclear fleet, as well as for safe shutdown of the reactor. When intermittent faults occur in cable circuits, the symptom of the problem often appears and disappears in milliseconds while the aftereffects can result in spurious alarms and automatic actuation of safety systems. Problems are compounded when component degradation causes a cable circuit to transmit inaccurate information that does not represent actual plant operating conditions. The root cause of intermittent cable circuit operation can include degradation of cable and connectors from heat, humidity, radiation, and vibration. Unlike hard faults (which are constant open or short circuits), intermittent faults are defined as a temporary change in impedance that may only last for a few milliseconds. While less disruptive than hard faults, signal spiking from intermittent faults can often lead to alarm conditions and reactor trips. A soft fault is defined as a subtle variation the electrical characteristics of the circuit that still permit the continued operation of the circuit. A soft fault may be a precursor to a hard fault due to continued degradation of the fault over time. The focus of this paper is the description of a cable testing system that can be connected to an active, low frequency circuit and continuously monitor electrical characteristics using the time domain reflectometry (TDR) technique until an intermittent impedance change occurs. This technique uses the difference in frequency response of the circuit and the TDR to isolate the TDR signature from the active circuit data . When a momentary impedance change occurs that exceeds a preconfigured limit threshold, the distance to fault data is automatically overlaid with the healthy baseline reference and stored in a database for diagnosis and repair. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Analysis and improvement of input channel selection algorithm for reliability of control systems in nuclear power plants - 138

In most of non-safety NSSS (Nuclear Steam Supply System) control systems, a process input is composed of two channels for redundancy. In a normal situation, the average of both input channels is used in the control algorithm. If a failed channel is detected, the 'channel selection algorithm' chooses the remaining normal input channel. For a bumpless transition, many implemented control systems include the 'ramping rate' function, which makes it possible for the selected input to slowly approach the new input channel. However, it recently turned out that this function adopted for positive effects may result in unintended consequences for the related control system and the plant in certain cases. Due to a process transient in the plant according to a significant event, any related sensor input signal can rapidly change and finally exceed the pre-defined range for detecting an out-of-range event in a short time. During the process of channel conversion due to a rapid transient in the plant, the processed input from the redundant channels can be distorted if the ramping rate function is not properly implemented. As a result, it can result in a plant shutdown because the control system fails to perform the critical function to mitigate a plant transient. To solve these potential problems, we analyzed the overall causes and derived a proposed algorithm for improving reliability and performance of the control systems. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

The overview in safety review of human factors engineering and control room design in chinese AP1000 nuclear power plant - 174

The first AP1000 nuclear power plant is constructed in Sanmen county in Zhejiang Province in China, which has creative and distinctive design characteristics. Human factors engineering disciplines are applied to the design of the AP1000. In addition to the elements of the program review model(including human factors engineering program management, operating experience review, functional requirements analysis and allocation, task analysis, staffing, integration of human reliability analysis with human factors engineering, human system interface design, procedure development, training program development, human system interface verification and validation, design implementation, human performance monitoring),the minimum inventory of controls, displays, and alarms present in the main control room and at the remote shutdown workstation. These contents mentioned above are reviewed according to standards, rules and regulations. This article introduces review process, and several important issues during the reviewing are described in detail. These issues include many elements, such as operating experience review, task analysis, human system interface design, verification and validation,and so on. This paper emphasizes on the main control room design(including environment and layout). Background noise in the actual main control room may exceed too much to design value. Another important issue is about main control room habitability systems(VES) changes to satisfy post-actuation performance requirements in AP1000 design change proposal. All the wall panel displays will be closed in this design change proposal. Review process and proposal are described in the paper. Since the AP1000 unit in China is the first constructed all over the world, verification and validation are especially important and necessary. The plan of verification and validation, the results summary report of verification and validation, the report of human engineering discrepancies are focus attention. Verification and validation in human factors engineering include HSI task support verification, HFE design verification and integrated system validation. This paper introduces the review process of verification and validation, the issues found in verification and validation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Online monitoring of rod control systems in combustion engineering pressurized water reactors - 195

Current designs of Pressurized Water Reactors (PWRs) are equipped with control and shutdown rods that are inserted and withdrawn from the reactor core to control reactivity. In Combustion Engineering (CE) plants, these rods are moved using the Control Element Drive Mechanism (CEDM) and controlled by the CEDM Control System (CEDMCS). In the event of system failure, troubleshooting activities can have a substantial impact on plant resources. Furthermore, CEDM component failures have occurred without warning, which has led to increased costs for the plants. These issues, combined with the inherently complex design of the CEDMCS, create a need for an online diagnostics system that can improve system health monitoring, increase diagnostic capabilities, and decrease troubleshooting time. This paper describes an advanced analysis and diagnostics system that monitors the CEDMCS during plant operation and provides critical information to onsite personnel to help detect and prevent problems before they impact normal plant operation. This system uses existing CEDMCS test points to acquire data and needs no modifications to the CEDMCS for implementation. The online monitoring capabilities automatically collect data during normal operation and diagnose failure events, reduce troubleshooting time and provide plant personnel with valuable information which is not currently available. Additionally, the system has built-in features used to streamline routine surveillance testing, collect stationary and movement data on multiple rods simultaneously, automate analysis, and generate reports. This system was developed with input from industry CEDMCS engineers at existing commercial facilities and has been deployed in multiple operating nuclear plants. The system has demonstrated itself to be a valuable tool for CEDMCS health monitoring, not only in reducing required man-power for routine system testing, but also in detected impending coil faults and allowing plant engineers to mitigate system component failures to prevent impending plant shut downs. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Virtual sensors for robust on-line monitoring (OLM) and diagnostics - 197

Unscheduled shutdown of nuclear power facilities for recalibration and replacement of faulty sensors can be expensive and disruptive to grid management. In this work, we present virtual (software) sensors that can replace a faulty physical sensor for a short duration thus allowing recalibration to be safely deferred to a later time. The virtual sensor model uses a Gaussian process model to process input data from redundant and other nearby sensors. Predicted data includes uncertainty bounds including spatial association uncertainty and measurement noise and error. Using data from an instrumented cooling water flow loop test bed, the virtual sensor model has predicted correct sensor measurements and the associated error corresponding to a faulty sensor. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Enhanced electrical penetration components meet the high safety standards required for China's HTR-PM Shidaowan and generation mPower small modular reactors - 300

Enhanced glass-to-metal feedthroughs will be used to carry key instrumentation and control (I and C) signals in China's high-temperature, gas-cooled pebble-bed modular reactor (HTR-PM) at Shidaowan, and in Small Modular Reactor (SMR) prototyping. Chinergy Co., Ltd and Jiamusi Electric Machine Co., Ltd will use the feedthroughs, also called electrical penetration assemblies (EPAs), at the Shidaowan twin-reactor high temperature reactor (HTR). These feedthroughs are manufactured using specialized processes that make them far more robust than counterparts made with organic or ceramic materials. Improved EPAs manufactured by SCHOTT are the only feedthrough solution that can be used in the primary loop of an HTR because organic seals cannot withstand the high temperatures and pressures of this environment. Glass-to-metal seals (GTMS) are made using inorganic, non-aging materials which enable a virtually unlimited lifespan in high-temperature, high-radiation and high-pressure applications. This I and C technology package is being enhanced by the development and testing of new mineral-insulated (MI) cable connectors and EPAs made of crystallized glass-ceramic combined with superalloys. The EPAs have a maintenance-free lifespan of over 60 years, preventing any outage time related to these components and subsequently reducing costs associated with shutdowns. They are far more robust than organic seals and provide increased safety at a lower lifetime cost. Resistance against high pressure and high temperature are especially ideal qualities that make EPAs a valuable choice for environments with high standards for fire protection, such as nuclear plants, subsea oil drilling, or any other underwater application. Due to their significant offering of safety-relevant qualities, the EPAs have also been chosen by BWXT for implementation in its mPower SMR. Small modular reactor requirements are complex and very challenging but are met by the new EPA components, which will help form a strong I and C safety chain. Unlike ordinary compression glass-to-metal seals, GTMS with crystallized glass-ceramic have the ability to withstand temperatures of over 350 deg. C / 662 deg. F and operating pressures of 3500 bar. Furthermore, their lifetime is qualified for up to 60 years, including pressure and radiation requirements in accordance with the latest IEC/IEEE standards. Higher-specification EPAs and MI cable connectors have been designed into severe accident I and C instruments, including advanced hydrogen and oxygen sensors that exceed the challenging requirements of both HTR and SMR applications. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Development of a transient signal validation technique via a modified kernel regression model - 365

The demand for robust and resilient performance has led to the use of online-monitoring and signal validation techniques to monitor the process parameters and equipment conditions using several empirical models. One of these models is kernel regression (KR). Although KR is being used and performed excellently when applied to steady state, it has limitation in time-varying data that has several repetition of the same data point of the signals, especially if those signals are used to infer the other signals. In addition, many situations are related to variation and fluctuation during normal states, such as transient - start-up and shutdown mode. However, accurate estimation of the process signal can lead to the proper understanding of the equipment behaviours as well as enhance the online monitoring applications. In the sight of foregoing, this paper proposed a modified KR model, especially for transient signal validation to resolve the setback of conventional KR by improving the measure of similarities in the conventional KR. The performance of the proposed model is verified using both experimental data and the plant simulation data, and its results are compared with that of conventional KR. The results show that the proposed model has a greater robustness in particular situations compared to conventional KR, and it can be used to improve process and equipment monitoring applications. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Development of a diversity and defense-in-depth strategy for the TerraPower TWR-P Advanced nuclear power plant - 380

A Diversity and Defense in Depth (D3) analysis was performed for the Traveling Wave Reactor, which is a Generation IV fast reactor design. The analysis demonstrated that the current I and C design for TWR-P provides sufficient diversity in the reactor shutdown system, the engineered safety features, and the post-accident monitoring system. In addition, an analysis was performed for the Embedded Digital Devices (EDD) to ensure a postulated Software Common Cause Failure (SWCCF) will not affect the safety functions of the TWR-P I and C system. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Regulatory Basis for Dry Spent Nuclear Fuel Storage Canister Inspections

Nuclear power plants in the United States began moving Spent Nuclear Fuel (SNF) from wet storage to dry storage in Independent Spent Fuel Storage Installations (ISFSIs) in 1986 at Surry station in Virginia. Dry storage represented an innovative solution to the problem created by the U.S. government's failure to fulfill its obligation to begin removing SNF from power plant sites. Today, about 70 ISFSIs have been placed into operation on the sites of current, or former nuclear power plants. There are currently over 2,200 SNF dry storage systems (DSSs) in service at these locations. Thirteen of these ISFSIs are located at sites with no operating reactor, and this number is due to increase in the coming years. Operating nuclear power stations use ISFSI storage to maintain wet storage capacity for recently discharged SNF. Permanently shutdown nuclear power plants use ISFSI storage to allow decommissioning and dismantlement of the power plant structures, systems, and components. The U.S. Nuclear Regulatory Commission (NRC) regulates the licensing, construction, and operation of ISFSIs under Title 10 of the U.S. Code of Federal Regulations, Energy, Part 72, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater than Class-C Waste (10 CFR 72). When the regulations in 10 CFR 72 were created around 1980, 20 years was chosen by the NRC as the license term for a specific ISFSI license and later, for a DSS Certificate of Compliance (CoC). Twenty years was, at the time, believed to be a sufficient time period for on-site interim dry storage of spent nuclear fuel before the U.S. government began moving the spent fuel away from the plant site for final disposal. Little, if any consideration was given to providing a regulatory framework for extending ISFSI licenses and DSS CoCs beyond 20 years at that time. However, shortly after the turn of the century, with still no federal program for removing SNF from reactor sites on the near-term horizon it became apparent that most DSSs would be required to remain in service at the ISFSIs beyond the initial 20-year license term. Operation beyond the original license term requires the specific license or the DSS CoC (used at a general license ISFSI) to be renewed. The requirements for renewal of a specific ISFSI license or a DSS CoC include addressing age-related degradation of structures, systems, and components that perform or support safety functions. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Development of a Discrete Event Simulation Model for Pyroprocessing Safeguardability

Advanced nuclear energy systems offer benefits for energy sustainability, independence, security, economic growth, and carbon emission reduction. Nuclear fuel fabricated for these systems utilize remotely-handled facilities for batch processing in hot cells. Because special nuclear material is in both different physical and chemical form than that of the contemporary fuel cycle, new challenges arise in terms of safeguardability. To that end, we are developing a system model for a commercial pyroprocessing facility that can also produce quantitative metrics to assess facility design. This paper presents the progress of a first build of a material throughput model using discrete event simulation for the fuel fabrication subsystem. The intent is to study the Type I error, or false alarm, as a metric for safeguardability. Initiating events that could cause false alarms arise from safeguards, safety, and security. False alarms can necessitate a facility shutdown until the anomaly is resolved by facility inspection. A high number would be costly for the facility operator both in terms of operation throughput goals and possible compensatory penalties. Therefore, different conceptual facility designs exhibit different frequencies for the false alarm per initiating event, and this can provide a measure of the safeguardability of each design. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

An IRSN Project For Experimental Data Acquisition In The Frame Of Criticality Safety And Reactor Physics

In the frame of its research missions for criticality safety assessment support, IRSN performed various experimental programs related to criticality safety in the CEA Valduc facilities. Following the shutdown of these facilities and the cancellation of the MIDAS project, which aimed at refurbishing these facilities, IRSN has initiated a new project named PRINCESS for 'Project for IRSN Neutron physics and Criticality Experimental data Supporting Safety'. PRINCESS is the practical implementation of IRSN's international partnership strategy to build relations with organizations that are equipped with experimental facilities for criticality safety and reactor physics research. The aim is to analyze experimental data from programs already performed and to conduct joint research programs on topics of interest of both parties. This paper presents an overview of the status of the PRINCESS project. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Consideration for Implementing the Safety Regulation Plans on Permanent Shut-own NPP during Transition Period in Korea

Recently the Korean regulatory body has been reviewing the change of permission on permanent shutdown Kori NPP (Nuclear Power Plant) unit 1, since the licensee (KHNP, Korea Hydro and Nuclear Power Co. Ltd) decided the permanent shut-down in June 2015 and applied the amendment of its licensing basis documents in June 2016. Although our institute with technical authority, the KINS (Korea Institute of Nuclear Safety), had performed the peer reviews for the construction permits of 3 NPPs and operating licenses of 25 NPPs, the safety control of permanent shut-down NPP and the transition activities including deactivation and decommissioning preparation are still in early stages and challenging fields. Currently the operation/only-possession system under defueled condition were defined through the peer review of fully revised documents and the major issues which concerned with the rearrangement of safety class or function based on new accident analysis results were produced. And also various transition activities were drawn as reflecting the decommissioning strategy on Kori unit 1. In this study, we described the status and problem of safety regulation based on our subordinate statute for permanent shut-down and transition activities. And then the detail implementation plans or the alternatives will be provided and discussed in order to solve the lack of regulatory experience and guide.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗