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853 records · Page 48
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Development of the B-25 Low Level Radioactive Waste Container in Taiwan - 1900x

Chinshan Nuclear Power Plant will be permanently stopped its operation in the middle of 2019 and will be the first nuclear power plant stepping into the decommissioning phase in Taiwan. According to Nuclear Reactor Facilities Regulation Act, a nuclear facility must be dismantled in 25 years after being permanently ceased operation. In its decommissioning plan, the B -25 low level radioactive waste container was one of the proposed containers used for this decommissioning task. However, the B-25 waste container was not used or developed in Taiwan before. To meet this need, Institute of Nuclear Energy Research developed a new type of B-25 waste container and will apply to the authority for its usage license. This article provided the information of the low level radioactive waste containers have been approved in Taiwan and described the design specifications of the developed B-25 waste container. Besides, the numerical evaluation and test results were also included in this article. Based on the evaluation and test results, the developed B-25 waste container satisfied the requirements of the storage container and industrial package type 2. It is believed that the usage license would be granted soon after applying it to the authority. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Unlocking the Vault - Accelerated Delivery Through the Application of Innovation and Collaboration - 19011

The Pile Fuel Cladding Silo (PFCS) at Sellafield was built in 1949/50 to receive and store fuel cladding from reprocessing of Windscale Pile fuel; it has been used to store this material and other waste from the UK nuclear programme ever since. The Silo is unsuitable for indefinite storage; this along with uncertainty in the make-up and condition of the inventory combine to rank the facility alongside the most hazardous in Western Europe. Sellafield Ltd, began working with its supply chain partners Bechtel Cavendish Nuclear Solutions (BCNS - a joint venture between Bechtel Inc and Cavendish Nuclear Ltd) in 2012, to execute a suite of projects to create an access route to the waste and a capability to retrieve waste from the silo for the first time ever. However, the challenge of providing this capability, while maintaining ongoing safe storage of the waste, and underpinning future performance of both the silo and the waste packages for many hundreds of years, proved to be too much for traditional Sellafield Ltd delivery models. With both cost and schedule forecasts doubling from initial estimates, a new approach was clearly required. The team recognised that success would require all of the experience and expertise of each organisation working together to a common goal. Therefore, significant effort was expended to build a collaborative team environment into which an innovative approach to both the engineering of what was provided and the delivery methods for doing it could be launched. Focussing on collaborative culture in which innovation and fit for purpose thinking were encouraged has completely rebuilt stakeholder confidence. The project is now almost complete; 17 months ahead of schedule and British Pounds 100 m less than the approved cost. (authors)

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DOE OSTI2019

Challenges of SLOWPOKE-2 HEU Irradiated Fuel Transportation from Canada to USA - 19123(2)

The SLOWPOKE-2 (Safe Low-Power Kritical Experiment) was designed by AECL (Atomic Energy of Canada Limited) in the late 1960's. The core consists of HEU (highly enriched uranium) fuel with an enrichment of 93%. The fuel cage typically has approximately 300 fuel pins assembled in a triangular pitch between the two plates. The fuel pins are made from uranium-aluminum alloy cladded with aluminum. The SLOWPOKE-2 reactor is rated at a nominal 20 kW. with varying neutron flux, maximum being 1 E12 n/cm{sup 2}/s. The mass of the fuel cage is 5.06 kg. The life of the core can be extended by adding layers of beryllium reflector. The transportation flask for the irradiated fuel is a Type B(U)F. The shipment classification is UN3328 Class 7. To date, 4 SLOWPOKE-2 reactors have been decommissioned. The fuel from all these reactors was transported to the U.S. The CNSC (Canadian Nuclear Safety Commission) joint Canada-United States guide for approval of Type B(U) and fissile material transportation packages (RD-364) satisfies the requirements stipulated in International Atomic energy Agency (IAEA) TS-R-1 1996 edition (revised) and is acceptable to the DOT (Department of Transportation) and NRC (Nuclear Regulatory Commission) for complying with U.S. code of federal regulations 10 CFR Part 71. The export license is issued by the CNSC. The transportation licence is issued by CNSC after approval from TC (Transport Canada) and also by DOT after approval from NRC. The final home of the HEU SLOWPOKE-2 irradiated fuel is in Savannah River Site of DOE (Department of Energy). The transportation licence requires that the Type B(U)F flask is structurally sound in case of a crash. The US DOT provides a CAC (Competent Authority Certificate) which describes the irradiated fuel specifications that can be transported in the designated flask. The U.S. DOT also requires a leak test certification in accordance with ANSI N14.5-2014 to verify a leakage rate not to exceed 1 E-7 std-cm{sup 3}/s. As most SLOWPOKE-2 facilities are not located on the existing approved routes for nuclear fuel transportation, a route survey needs to be performed and submitted to the NRC. The route survey report is assessed by the NRC and issues a route approval for the transportation of the fissile material. This technical paper will discuss the detailed processes required to obtain approvals in order to transport the HEU SLOWPOKE-2 irradiated fuel from Canada to U.S. based upon recent experiences of decommissioning SLOWPOKE-2 reactors in Canada. It will provide details on the requirements of various regulators in Canada and the U.S. that govern and regulate this shipment including transportation, safety and security. It will also present lessons learned from previous experiences that can be utilized for best practice on other similar projects. (authors)

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DOE OSTI2019

Contamination Control: Challenges and Innovative Design Solutions utilized at Chalk River Laboratories - 19263

CNL has operated research reactors at the Chalk River Laboratories (CRL) site since 1947, generating a variety of spent nuclear fuel wastes produced by irradiating metallic uranium and other prototype fuels. This legacy waste has been stored in below-grade, vertical cylindrical steel and concrete structures known as Tile Holes. Ongoing monitoring of the legacy waste has identified moisture resulting from incomplete drying prior to emplacement, surface water and humidity. The undesirable storage conditions result in corrosion of both the legacy waste and the storage containers. Continued storage in these conditions may result in further degradation of the storage container and contamination of the surrounding environment. In 2004, the Fuel Packaging and Storage (FPS) Project was initiated, with the objective to develop a complete system to safely retrieve legacy waste from these Tile Holes as well as provide a stable storage environment until a final long-term management plan is established. Contamination control during retrieval, transfer and repackaging operations is critical for personnel and environmental safety. The significant challenges with retrieval of the legacy waste from the Tile Holes includes management of loose collateral liquid and solids that will be brought into the system, with the retrieved waste container. During the retrieval, transfer and repackaging operation, the challenge was to: 1) Prevent an environmental release by potentially transferring these contaminants onto the external surface of the Tile Hole Array; 2) Prevent an environmental release and ensure personnel safety while transporting the legacy waste; 3) Control release of contaminants into the new clean storage container while preventing contamination of the packaging equipment. To address the first source of contamination, a debris guide liner capable of interfacing with multiple Tile Hole configurations was integrated within the Retrieval Flask to limit contamination of the flask interior and to direct collateral liquids and solids back into the Tile Hole. The system uses a multi-function sealed Shield Gate and Drip Tray to retain and prevent the contaminants from escaping while transferring the legacy waste. This gate, along with a thin walled, Disposable Debris Guide Liner (DDGL) is an innovative solution developed to control the flow characteristics of the residual waste and to contain the waste to protect the internal components of the repackaging equipment. The development of the complete solution was further challenged as the DDGL must not tear or rip and must compact inside the new storage containers for a range of legacy waste storage containers. (authors)

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DOE OSTI2019

Career Development through Discreet Simulation Modeling - 19009

The cleanup mission of the Department of Energy (DOE) Office of River Protection (ORP) at the Hanford Site is to immobilize 212,000 m3 (56 million gallons) of nuclear waste, currently stored in 177 underground tanks, at the Waste Treatment and Immobilization Plant (WTP) and associated supplemental treatment facilities. Waste transfers to and from the WTP utilize the intricate Hanford Waste Feed Delivery (WFD) system, a network composed of: double-shell tanks (DSTs), single shell tanks (SSTs), the effluent treatment facility (ETF), a large number of pumps, valves, and leak detectors, and thousands of feet of underground transfer lines. This equipment must be prepared to safely and reliably transfer waste to the WTP. This study used discrete event simulation modeling to explore the behavior of the waste transfer system to identify areas of improvement. Two models and their subsequent results are presented: the Effluent Treatment Facility and the next-to-be-retrieved single-shell tanks (A/AX Farms). This study focused on identifying throughput bottlenecks, operational constraints, choke-point analysis, and redundancy in equipment, process design, and operational strategy. Analysis performed on the ETF OR Model indicated that the existing facility operation strategy will not result in a sufficient throughput to prevent WTP melter idling. Furthermore, in the absence of equipment failures, the facility was still shown to be unable to reach the required throughput to prevent melter idling. It is concluded that ETF system maintenance events are currently the main bottlenecks in the ETF facility. Analysis performed on the SST (A/AX) OR model indicate that the main drivers of the retrieval schedule are the hard retrieval start dates for each SST tank. In addition, it was concluded that operating the ventilation trains in parallel results in 15 times less retrieval interruptions and shortens the retrieval durations. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Waste Management Analyses for Spent Structures at SNS - Paper 106

The Spallation Neutron Source (SNS) in Oak Ridge, Tennessee, is an accelerator driven neutron scattering facility for materials research. Presently, the SNS operates at 1.2 megawatt (MW) proton beam power incident on a mercury target with a proton beam energy of 1 GeV and 60 Hz repetition rate. The SNS consists of an accelerator facility, target facility, and a world-class suite of neutron scattering instruments supporting material, life-science and fundamental physics research. According to the SNS operations plan, some components are replaced when they reach their end-of-life due to radiation induced material damage or burn-up, or because of mechanical failure. During operation, these components, especially those in proximity to the target, are exposed to a radiation environment and build up significant activity during their service lifetime. All these components must be safely removed, placed in a container/package for storage, and ultimately transported offsite for disposal. An accurate estimate of the radionuclide inventory is performed assuming realistic irradiation history and decay time in order to classify a spent component. On the basis of a calculated radionuclide inventory, the spent component is classified, and an appropriate cask/package for transport and storage is suggested. After the choice for the container is made, radiation transport calculations are performed for the facility component placed inside the container, to ensure that the container is compliant with waste management regulations. When necessary, additional shielding is added. For each spent component before it goes out of the SNS site, supporting documentation is produced with a radionuclide inventory prediction, dose rate analyses and, if required, additional shielding. A dose rate analyses for the exposure prediction of personnel during components change out are performed as well. Most of the effort is concentrated on the target facility components such as target, proton beam window (PBW) and core vessel insert (CVI) plugs because of severe radiation conditions. Additionally, analyses are performed for the accelerator facility components, which are removed from the facility. Activation analyses and residual dose rates analyses are performed to support safe handling of SNS spent components at the SNS site and during transport to the waste disposal site. On the basis of the calculated radionuclide inventories, spent components characterized and classified with respect to waste management guidelines, and activity-based determinations are made regarding appropriate packaging for transport. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2015

MONK10: Burnup credit capability - 14596

MONK{sup R} is a Monte Carlo code for nuclear criticality and reactor physics analyses. It has a proven track record of application to the whole of the nuclear fuel cycle and is well established in the UK criticality community. Furthermore it is increasingly being used for reactor physics analysis (as described at ICNC 2011), which makes it an ideal tool for burn-up credit (BUC) calculations. Throughout the paper, example calculations based on a PWR are presented to illustrate the capabilities of the MONK10 code. In order to account for the spatial dependence of material burn-up it has in the past been necessary to design models with multiple regions and materials specifically to allow material burn-up to vary spatially. This is very labour intensive and difficult to change at a later stage. A new code version, MONK10, was released last year which includes the facility to allow a burn-up (BU) mesh to be superimposed on an existing model in order to account for the spatial dependence of the burn-up. This facility is used to consider the effect of radial position of a fuel element in a PWR core on BUC. Additionally, a thermal hydraulics (TH) mesh can be used to specify region dependent temperature. This, coupled with the fact that MONK10 also incorporates an on-the-fly Doppler broadening methodology facilitates the modelling of spatially dependent temperatures for the different components. A TH mesh is used to superimpose a temperature profile on a PWR based model and the effect of this on BUC is considered. The burn-up modelling in MONK has been benchmarked against the ANSWERS WIMS deterministic reactor physics code. Once the burn-up calculation has been completed and the depleted fuel compositions determined the spent fuel compositions can be transferred into a model of a storage facility or transport flask in order to perform burn-up credit analysis. The initial model is usually described as the donor model and the latter model as the receiver model. This transfer is carried out using the COWL option which allows the specification of a material in the receiver model based on the material's composition in a given BU mesh cell from the donor model. This allows compositions and densities to be transferred and also allows user specified adjustments to be made. For example, this could include omitting the fission products in order to estimate their contribution to burn-up credit and provide an actinide-only analysis. The effect of excluding appropriate nuclides is presented. An example of how the ANSWERS SPRUCE code can be used to quantify uncertainty in a BUC calculation is also presented. (authors)

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