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At least 523 records · Page 29

AIR SHIPMENT OF SPENT NUCLEAR FUEL FROM THE BUDAPEST RESEARCH REACTOR

The shipment of spent nuclear fuel is usually done by a combination of rail, road or sea, as the high activity of the SNF needs heavy shielding. Air shipment has advantages, e.g. it is much faster than any other shipment and therefore minimizes the transit time as well as attention of the public. Up to now only very few and very special SNF shipments were done by air, as the available container (TUK6) had a very limited capacity. Recently Sosny developed a Type C overpack, the TUK-145/C, compliant with IAEA Standard TS-R-1 for the VPVR/M type Skoda container. The TUK-145/C was first used in Vietnam in July 2013 for a single cask. In October and November 2013 a total of six casks were successfully shipped from Hungary in three air shipments using the TUK-145/C. The present paper describes the details of these shipments and formulates the lessons learned.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

SRNL Development of Recovery Processes for Mark-18A Heavy Actinide Targets

Savannah River National Laboratory (SRNL) and Oak Ridge National Laboratory (ORNL) are developing plans for the recovery of rare and unique isotopes contained within heavy-actinide target assemblies, specifically the Mark-18A. Mark-18A assemblies were irradiated in Savannah River Site (SRS) reactors in the 1970s under extremely high neutron-flux conditions and produced, virtually, the world's supply of plutonium-244, an isotope of key importance to high-precision actinide measurement and other scientific and nonproliferation uses; and curium highly enriched in heavy isotopes (e.g., curium-246 and curium-248). In 2015 and 2016, SRNL is pursuing tasks that would reduce program risk and budget requirements, including further characterization of unprocessed targets; engineering studies for the use of the SRNL Shielded Cells Facility (SCF) for recovery; and development of onsite and offsite shipping methods including a replacement for the heavy (70 ton) cask previously used for onsite transfer of irradiated items at SRS. A status update is provided for the characterization, including modeling using the Monte Carlo N-Particle Transport Code (MCNP); direct non-destructive assay measurements; and cask design.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

Oak Ridge National Laboratory (ORNL) and the US Nuclear Regulatory Commission (NRC) have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation systems (often referred to as casks) and spent fuel pools (SFPs). This work is divided into two main phases. The first phase investigated the applicability of peak reactivity methods currently used in SFPs to transportation and storage casks and the validation of reactivity calculations and spent fuel compositions within these methods. The second phase focuses on extending BUC beyond peak reactivity. This paper documents the analysis of the effects of control blade insertion history, and moderator density and burnup axial profiles for extended BWR BUC.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Systems and methods for harvesting and storing materials produced in a nuclear reactor

Systems produce desired isotopes through irradiation in nuclear reactor instrumentation tubes and deposit the same in a robust facility for immediate shipping, handling, and/or consumption. Irradiation targets are inserted and removed through inaccessible areas without plant shutdown and placed in the harvesting facility, such as a plurality of sealable and shipping-safe casks and/or canisters. Systems may connect various structures in a sealed manner to avoid release of dangerous or unwanted matter throughout the nuclear plant, and/or systems may also automatically decontaminate materials to be released. Useable casks or canisters can include plural barriers for containment that are temporarily and selectively removable with specially-configured paths inserted therein. Penetrations in the facilities may limit waste or pneumatic gas escape and allow the same to be removed from the systems without over-pressurization or leakage. Methods include processing irradiation targets through such systems and securely delivering them in such harvesting facilities.

07 ISOTOPE AND RADIATION SOURCES↗

Experimental Test Plan for PWR Sister Rods in the High Burnup Spent Fuel Data Project

The High Burnup Spent Fuel Data project pulled 25 sister rods (9 from the project assemblies and 16 from similar HBU assemblies) for characterization. The 25 sister rods are all high burnup and cover the range of modern domestic cladding alloys. The 25 sister rods were shipped to Oak Ridge National Laboratory (ORNL) in early 2016 for detailed non-destructive and destructive examination. Examinations are intended to provide baseline data on the initial physical state of the cladding and fuel prior to the loading, drying, and long-term dry storage process. Further examinations are focused on determining the effects of temperatures encountered during and following drying. Similar tests will be performed on rods taken from the project assemblies at the end of their long-term storage in a TN-32 dry storage cask (the cask rods ) to identify any significant changes in the fuel rods that may have occurred during the dry storage period. Additionally, some of the sister rods will be used for separate effects testing to expand the applicability of the project data to the fleet, and to address some of the data-related gaps associated with extended storage and subsequent transportation of high burnup fuel. A draft test plan is being developed that describes the experimental work to be conducted on the sister rods. This paper summarizes the draft test plan and necessary coordination activities for the multi-year experimental program to supply data relevant to the assessment of the safety of long-term storage followed by transportation of high burnup spent fuel.

dry storage↗

Extending Spent Fuel Storage until Transport for Reprocessing or Disposal

Spent fuel (SF) must be stored until an end point such as reprocessing or geologic disposal is imple-mented. Selection and implementation of an end point for SF depends upon future funding, legisla-tion, licensing and other factors that cannot be predicted with certainty. Past presumptions related to the availability of an end point have often been wrong and resulted in missed opportunities for properly informing spent fuel management policies and strategies. For example, dry cask storage systems were originally conceived to free up needed space in reactor spent fuel pools and also to provide SFS of up to 20 years until reprocessing and/or deep geological disposal became available. Hundreds of dry cask storage systems are now employed throughout the world and will be relied upon well beyond the originally envisioned design life. Given present and projected rates for the use of nuclear power coupled with projections for SF repro-cessing and disposal capacities, one concludes that SF storage will be prolonged, potentially for several decades. The US Nuclear Regulatory Commission has recently considered 300 years of storage to be appropriate for the characterization and prediction of ageing effects and ageing management issues associated with extending SF storage and subsequent transport. This paper encourages addressing the uncertainty associated with the duration of SF storage by de-sign – rather than by default. It suggests ways that this uncertainty may be considered in design, li-censing, policy, and strategy decisions and proposes a framework for safely extending spent fuel storage until SF can be transported for reprocessing or disposal – regardless of how long that may be. The paper however is not intended to either encourage or facilitate needlessly extending spent fuel storage durations. Its intent is to ensure a design and safety basis with sufficient margin to accommodate the full range of potential future scenarios. Although the focus is primarily on storage of SF from commercial operation, the principles described are equally applicable to SF from research and production reactors as well as high-level radioactive waste.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

A Burnup Credit Approach for Margin Estimation of Loaded Boiling Water Reactor Canisters in UNF-ST&DARDS

This paper discusses the modeling approach applied in the Used Nuclear Fuel Storage, Transportation, & Disposal Analysis Resource and Data System (UNF-ST&DARDS) to take credit for the reduced reactivity associated with the burnup of boiling water reactor (BWR) fuel for canister-specific as loaded criticality analysis. Burnup credit is routinely used for the analysis of pressurized water reactor (PWR) spent nuclear fuel (SNF) storage (wet), transportation, and disposal. The fresh fuel assumption has historically been applied to BWR SNF criticality analyses for dry storage and transportation, and the peak reactivity method of burnup credit has been applied to pool storage. The fresh fuel assumption ignores the credit for gadolinium burnable absorbers in BWR fuel, as well as the burnup of the fuel, and is therefore conservative. This work employs the framework traditionally used for PWR burnup credit analyses and applies available data and recently published regulatory guidance to justify a new approach to model the as-loaded burnup and enrichment of SNF assemblies in currently loaded dry casks. The following features of this criticality analysis approach are presented in this paper: (1) selection of axial burnup profiles for BWR fuel from publicly available sources using methods derived from recently published research, (2) justification of the selected axial burnup profiles as relevant to all types of BWR fuel, (3) justification for modeling the fuel assemblies with a uniform axial and radial enrichment, and (4) justification for modeling the axial void profile and control blade insertion during depletion. This analysis approach was used to evaluate 153 already-loaded BWR casks at 8 sites as a function of time in UNF ST&DARDS. Finally, the results of the as-loaded criticality safety margin assessments are presented. The results show the k eff calculated for storage and transportation cases range from 0.71 to 0.83, and the eigenvalues calculated for the disposal cases range from 0.85 to 1.03, though the highest cases are dominated by very limiting hypothetical damaged fuel assumptions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Planning for Receipt of Mk-18A Recovered Nuclear Material at Oak Ridge National Laboratory

Oak Ridge National Laboratory (ORNL) ships and receives a wide variety of radioisotopes. These shipments will include the Mark-18A high-burnup targets, which are being processed at Savannah River National Laboratory to recover valuable isotopes that will then be shipped to ORNL for programmatic use. The Mark-18A targets will be processed to recover two products: a plutonium oxide stream and an oxide material that will contain the americium, curium, and lanthanides. A Type A shipping package has been designed, fabricated, and certified to transport the americium/curium/lanthanide oxide product in special form capsules. The package was designed with both gamma and neutron shielding for contents reading up to 5,000 R/hour. The plutonium stream will be shipped to ORNL in a Model 9975 Type B transport cask. A new content envelope for the Type B cask is being developed. This paper discusses the preparation underway at ORNL to receive the packages and transfer the material into hot cell processing and storage facilities for subsequent use in research and development applications and as feedstock for heavy actinide production.

Patton, Bradley↗

Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites

In this paper, a preliminary evaluation of removing used nuclear fuel (UNF) from 12 shutdown sites was conducted. Shutdown sites are defined as those commercial nuclear power reactor sites where the nuclear power reactors have been shut down and the site has been decommissioned or is undergoing decommissioning. The shutdown sites were Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, and San Onofre. The evaluation was divided into four components: • characterization of the UNF and greater-than-Class C low-level radioactive waste (GTCC waste) inventory • a description of the on-site infrastructure and conditions relevant to transportation activities • an evaluation of the near-site transportation infrastructure and experience relevant to shipping transportation casks containing UNF from the shutdown sites, including gaps in information • an evaluation of the actions necessary to prepare for and remove UNF and GTCC waste from the shutdown sites. The primary sources for the inventory of UNF and GTCC waste are the DOE RW-859 used nuclear fuel inventory database, industry sources such as StoreFUEL and SpentFUEL, and government sources such as the NRC. The primary sources for the information on the site conditions and near-site transportation infrastructure and experience include site visits to the Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, and Zion sites; information provided by managers at the shutdown sites; Facility Interface Data Sheets compiled for DOE in 2005; Services Planning Documents prepared for DOE in 1993 and 1994; industry publications such as Radwaste Solutions; and Google Earth. State and Regional Group representatives, a Tribal representative, and a Federal Railroad Administration representative participated in six of the shutdown site visits. All sites were found to have at least one off-site transportation mode option for removing their UNF and GTCC waste, and some sites have two options. Large component removals during reactor decommissioning provided an important source of information in determining transportation mode options. Especially important in conducting the evaluation were site visits, through which information was obtained that would not have been available otherwise. Extensive photographs taken during the site visits proved to be particularly useful in documenting the current conditions at or near the sites. Additional conclusions from this evaluation include: • Although there are common aspects, each site has some unique features and/or conditions. • The 12 sites use designs from 4 different suppliers, including 9 different (horizontal and vertical) storage systems that would require 8 different transportation cask designs. • Although some regulatory actions will be required, all UNF at the initial 9 shutdown sites (Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, and Zion) is in licensed systems that can be transported, including a small amount of high-burnup fuel. • Each site indicated that 2-3 years of advanced time is required for preparations prior to pick-up. • Most sites have multiple transportation options, e.g., rail, barge, or heavy haul truck, as well as constraints and preferences. It is expected that additional site visits will be conducted to update the information used in the evaluation. Additional sites may be added to the evaluation if further site closures are announced.

shutdown sites, used nuclear fuel, NFST, Nuclear F↗

International MNSR Training Center in Ghana Provides HEU Removal Expertise

Building on our years of cooperation, the United States and Ghana worked together to construct the International MNSR Training Center (IMTC) at the National Nuclear Research Institute of the Ghana Atomic Energy Commission (NNRI/GAEC) in Accra, Ghana. Here, the IMTC was used for training operators in Ghana to discharge the reactor’s HEU core, load it into a shielded transportation cask, and secure the cask for transport. Moving forward, the IMTC will be used to train personnel from the remaining countries with MNSRs to conduct these operations.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Self-powered Wireless Through-wall Data Communication for Nuclear Environments

In the nuclear industry, many vital components, such as spent fuel storage canisters and nuclear reactor pressure vessels (RPV), are entirely enclosed by metal and surrounded by thick concrete walls that manage the potentially harmful radiation and prevent release to the environment. Due to the casks’ long storage, monitoring temperature, pressure, radiation, humidity, structural health, etc., within these enclosed vessels is crucial to ensure the fuel containment safety and security. In this paper, a self-powered wireless through-wall data communication system for nuclear environments was designed. The whole package includes: a radiation energy harvester with power management; ultrasound wireless communication using high-temperature piezoelectric transducers; electronics modules for harvesting, sensing, and data transmission; and radiation shielding for electronics and sensors. The package is able to harvest tens to hundreds of mW power from a nuclear canister environment directly and provides a path forward for continuous monitoring of the dry cask for 50 years.

Wu, Yongjia↗

Damaged Fuel in the United States

Over the years of commercial nuclear power plant operations in the United States (U.S.), many fuel assemblies have lost the capability to perform all of their desired functions. These assemblies have lost the capability to be handled, stored, or transported to meet regulations resulting in them being classified as damaged fuel. The causes of these failed assemblies are diverse and plant-specific. The majority of these failed assemblies are contained in a damaged fuel can that will be used in conjunction with a storage and/or transportation system. The storage and transportation systems have a limit on how many slots can be filled with damaged fuel cans. A damaged fuel can is generally a stainless steel container that confines damaged SNF and is closed on its end by screened openings. These screened openings allow gaseous and liquid media to escape but minimize the dispersal of gross particulate material. Out of an abundance of caution a few reactors have loaded high burnup fuel into damaged fuel cans. Damaged fuel is not licensed for storage or transport in the U.S., because the regulations for storage and transport do not specify exactly how to classify damaged fuel. Instead the regulations license/certify packages that specify approved contents. Damaged fuel must be included in the approved contents to be a viable option. In many cases damaged SNF is encapsulated in a damaged fuel can to ensure it can confine gross fuel particles, debris, and or damaged assemblies to a known volume within a loaded cask. This damaged fuel can may then be utilized in the same way as an assembly in a storage and transportation design system. Some storage cask systems utilize top and bottom plugs to confine debris in damaged fuel. The most recent domestic documentation on damaged SNF was performed by the U.S. Energy Information Administration (EIA) and used data from U.S. reactors concluding in June 30th 2013 to produce Form GC-859, “Nuclear Fuel Data Survey”. Based on this form there were 136,821 boiling water reactor (BWR) spent nuclear fuel (SNF) assemblies and 104,647 pressurized water reactor (PWR) SNF assemblies for a total of 241,468 SNF assemblies in the U.S. Out of these 241,468 SNF assemblies, 4,521 assemblies were classified as failed. Some assemblies were also disassembled and the fuel rods or pieces of fuel rods were combined together to make a consolidated assembly. A consolidated assembly may include damaged fuel, or it could have been consolidated as part of a demonstration project. The GC-859 data includes 2,550 consolidated assemblies containing between 0 and 264 whole fuel rods. These consolidated assemblies could be placed in single assembly canisters and stored in the spent fuel pool. For dry storage and transportation, a single assembly canister is generally placed in a damaged fuel can. In addition to the consolidated assemblies, 2,391 uncanistered fuel rod pieces exist, which were removed from 494 assemblies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

System Analysis Modeling and Intermodal Transportation for Commercial Spent Nuclear Fuel

The United States Department of Energy (DOE) has long term goals to develop solutions for managing the nation’s spent nuclear fuel (SNF) and high-level waste (HLW) inventory. The Integrated Waste Management (IWM) program under the DOE office of Nuclear Energy (DOE-NE) is employing system-level engineering and analysis principles to inform potential future waste management system architectures. Managing the spent nuclear waste requires the use of system-level analysis software that takes various aspects of the fuel cycle into account like waste generation, on-site/centralized storage, transportation infrastructure, and long-term disposal. The Next Generation System Analysis Model (NGSAM) is an agent-based model that was developed to simulate the transportation and storage of SNF and HLW. As an agent-based model, NGSAM has the capability to detail the interaction and movement of individual components and groups, such as rail cars and casks. The SNF inventory from commercial nuclear reactors is currently in temporary storage at multiple locations spread across the United States. Shipping of SNF from these locations relies on one of three transportation modes: rail, heavy-haul truck, or barge. Out of the three modes identified, rail is generally the most preferred due to the size of the canisters and casks the SNF would be shipped in. However, under some scenarios, a direct rail route might not be readily available to a reactor site or improving the rail infrastructure at shutdown sites might be too cost-prohibitive for utilities to opt for a direct rail transfer. Under such scenarios, using a barge or heavy haul truck to de-inventory the site and transfer the SNF to a nearby intermodal transfer site with adequate rail infrastructure where the payload could be transferred to a rail car might prove to be an attractive option. This work initially presents the various intermodal transportation options that could be used to transfer SNF from reactor sites to rail cars. This is followed by exploring the operational steps in each of these modes to move the SNF from a reactor site and transfer it to a rail car. This work also presents the procedure of implementing the intermodal transfer methodology in NGSAM using various Java methods. Finally, the process times for accomplishing each of the individual steps are furnished. The implementation ideology, assumptions, and future steps are presented in this work.

Gadey, Harish Reddy↗

THE SHOCK AND VIBRATION ENVIRONMENT FOR USED NUCLEAR FUEL TRANSPORTATION MODELING

Pacific Northwest National Laboratory (PNNL) is working on modeling the shock and vibration environment of used nuclear fuel (UNF) under normal conditions of transport for the Department of Energy (DOE). A cask and cradle system, loaded with surrogate fuel assemblies, experienced several different modes of transportation. During transport, strains and accelerations were recorded at various places in the cask and cradle system. These measurements were then analyzed and used to characterize the shock and vibration environment for future use in modeling normal conditions of transport. Overall, the strains that the fuel rods in the surrogate fuel assemblies experienced were low; however, the characterization of the shock and vibration environment provided valuable information for further modeling and analysis.

Ivanusa, Pavlo↗

Dry Storage of Research Reactor Spent Nuclear Fuel - 13321

Spent fuel from domestic and foreign research reactors is received and stored at the Savannah River Site's L Area Material Storage (L Basin) Facility. This DOE-owned fuel consists primarily of highly enriched uranium in metal, oxide or silicide form with aluminum cladding. Upon receipt, the fuel is unloaded and transferred to basin storage awaiting final disposition. Disposition alternatives include processing via the site's H Canyon facility for uranium recovery, or packaging and shipment of the spent fuel to a waste repository. A program has been developed to provide a phased approach for dry storage of the L Basin fuel. The initial phase of the dry storage program will demonstrate loading, drying, and storage of fuel in twelve instrumented canisters to assess fuel performance. After closure, the loaded canisters are transferred to pad-mounted concrete overpacks, similar to those used for dry storage of commercial fuel. Unlike commercial spent fuel, however, the DOE fuel has high enrichment, very low to high burnup, and low decay heat. The aluminum cladding presents unique challenges due to the presence of an oxide layer that forms on the cladding surface, and corrosion degradation resulting from prolonged wet storage. The removal of free and bound water is essential to the prevention of fuel corrosion and radiolytic generation of hydrogen. The demonstration will validate models predicting pressure, temperature, gas generation, and corrosion performance, provide an engineering scale demonstration of fuel handling, drying, leak testing, and canister backfill operations, and establish 'road-ready' storage of fuel that is suitable for offsite repository shipment or retrievable for onsite processing. Implementation of the Phase I demonstration can be completed within three years. Phases II and III, leading to the de-inventory of L Basin, would require an additional 750 canisters and 6-12 years to complete. Transfer of the fuel from basin storage to dry storage requires integration with current facility operations, and selection of equipment that will allow safe operation within the constraints of existing facility conditions. Examples of such constraints that are evaluated and addressed by the dry storage program include limited basin depth, varying fuel lengths up to 4 m, (13 ft), fissile loading limits, canister closure design, post-load drying and closure of the canisters, instrument selection and installation, and movement of the canisters to storage casks. The initial pilot phase restricts the fuels to shorter length fuels that can be loaded to the canister directly underwater; subsequent phases will require use of a shielded transfer system. Removal of the canister from the basin, followed by drying, inerting, closure of the canister, and transfer of the canister to the storage cask are completed with remotely operated equipment and appropriate shielding to reduce personnel radiation exposure. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Going for complete defueling - A comprehensive solution for damaged spent fuel

The complete defueling of a NPP is a prerequisite for decommissioning and dismantling. In particular the defueling of damaged spent fuel and the following dry storage remains a challenge. It is the aim of the Quiver project to develop a one-fits-it-all solution for all different kinds of damaged spent fuel and to develop in addition comprehensive handling equipment to ensure a thoroughly dispatch of the damaged spent fuel inside the Quiver. As a comprehensive solution, GNS developed a Quiver System which complies with the requirements of the transport and storage cask CASTOR V. This paper details, first, the loading of damaged fuel rods into the Quivers including the necessary equipment, secondly, the dispatch of the Quivers including an illustration of the required equipment and facilities, and thirdly, the loading and dispatch of Quivers in the CASTOR V casks.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Developments in the Regulatory Framework of Interim Storage of Spent Nuclear Fuel (SNF) in Germany - 17374

According to article 6 of the German Atomic Energy Act, storage of nuclear fuel requires a license, granted by the competent authority in this field, which has been the Federal Office of Radiation Protection (BfS) until July 2016 and currently is the Federal Office for the Safety of Nuclear Waste Management (BfE). A first license was granted in 1983 for storage of spent nuclear fuel (SNF) and vitrified high level waste (HAW) at the interim storage facility in Gorleben. This license was reviewed in 1995. Dry Interim storage at the NPP sites in its current form started in 2002 in the interim storage facility next to the NPP Lingen. Since that time at each site of an operational NPP storage facilities were licensed and erected. Thus additionally to the three central storage facilities (Ahaus, Gorleben and Rubenow) there are now 12 storage facilities onsite the NPPs. All of these facilities use dry storage in metallic dual purpose casks (DPCs). The actual storage licenses for SNF and the HAW (which is stored in Gorleben and Rubenow) are limited to 40 years after closing the transport and storage cask. The first license to expire will be the storage license for the interim storage facility in Gorleben in the year 2034. The German 'Programme for the safe and responsible management of spent fuel and radioactive waste' expects to take the disposal facility into operation in the year 2050. Thus necessary preparations for interim storage till that time have to be taken on the side of the regulator. In the German Atomic Energy Act precautions have already been taken and a consulting of the German parliament (Bundestag) is mandatory if inevitable events would require an extension of currently licensed interim storage period. In the year 2016, the regulatory body in Germany was given a new structure. The BfS was formerly competent authority in the field of radiation protection since 1989. It had competencies for ionizing and non-ionizing radiation protection (e.g. mobile phone radiation, high voltage power lines and the effects of electric fields), licensing storage of spent nuclear fuel and as the operator of facilities for final disposal. Now there is a new structure with two Federal Offices (BfS and BfE) and a federal company for final disposal (BGE). The BfE is responsible for regulation in the field of nuclear disposal since 30.07.2016. So the package design approval for DPCs and the storage licenses for SNF are now granted by this office. The search for a repository site will be coordinated there as well. The site selection, construction and operation of the repository will be in the responsibility of BGE. The parts of the former BfS dealing with recovery of the waste from Asse (a former salt mine used for disposal research) and the construction of the repository Konrad (for low and medium level radioactive waste) will be transferred into that company. It is planned to incorporate the German company for construction and management of waste disposal facilities (DBE) which is owned by the German utilities as well. The BfS will focus on the scientific foundations and further advancement of radiation protection. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Computational Safeguards Analysis of PWR Spent Nuclear Fuel in Dry Storage - 17435

The master plan for spent nuclear fuel (SNF) management in Korea recently adopted by Korean atomic energy promotion council. According to the master plan, the detailed practical and R and D strategies for SNF transferring, storage and disposal should be determined in the near future. Besides, it is also necessary to make plan how to secure SNF at each step in terms of nuclear safeguards, safety, security and nonproliferation. Several nondestructive analysis (NDA) techniques have been studied for assuring the safety of SNF in PWR methods. Obviously, gamma-ray is most dominant radiation emitted by SNF so several gamma-ray measurement methods have been developed and applied for the purpose of SNF monitoring. Those methods, however, suffer from a lack of nuclear data and interfering background radiation. To date, several neutron measurement methods have been developed as a solution to the problem of gamma-ray methods. Korea Institute of Nuclear Nonproliferation and Control (KINAC) currently develops a new novel technique of fast neutron detection for spent fuel monitoring at dry spent fuel storage facilities. As the first step, a computational program has been developed in order to evaluate neutron flux on the surface of PWR dry storage cask and study for some potential sabotage events. Typically, neutron flux on the surface of dry storage cask can be evaluated via two steps. The source term, based on burn-up rate of SNF, its cooling time and others related to operation condition at nuclear power plants, can be firstly defined by ORIGEN code. The surface flux can be then evaluated via MCNP code. These two separated steps are complicated and time consuming so KINAC developed a new algorithm and user-interface, coupled ORIGEN and MCNP. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗