scholarly journals Evaluation of alternative review plans for the Department of Energy safety analysis and review system

1981 ◽  
Author(s):  
E.T. Browne ◽  
A.F. Jr. Meyer
Keyword(s):  
Safety Analysis ◽  
Department Of Energy ◽  
Review System

Demonstration and Licensing of a Spent Nuclear Fuel Dry Transfer System

2001 ◽  
Author(s):  
Mikal A. McKinnon ◽  
Leroy Stewart
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Safety Analysis ◽  
The United States ◽  
Lessons Learned ◽  
Department Of Energy ◽  
Transfer System ◽  
National Academy Of Sciences ◽  
Cooperative Effort ◽  
Site Specific

Abstract Research studies by the Electric Power Research Institute (EPRI) established the technical and operational requirements necessary to enable the onsite cask-to-cask dry transfer of spent nuclear fuel. Use of the dry transfer system has the potential to permit shutdown reactor sites to decommission pools and provide the capability of transferring assemblies from storage casks or small transportation casks to sealed transportable canisters. Following an evaluation by the Department of Energy (DOE) and the National Academy of Sciences, a cooperative program was established between DOE and EPRI, which led to the cost-shared design of a dry transfer system (DTS). EPRI used Transnuclear, Inc., of Hawthorne, New York, to design the DTS in accordance with the technical and quality assurance requirements of the code of Federal Regulations, Title 10, Part 72 (10CFR72). EPRI delivered the final design report to DOE in 1995 and the DTS topical safety analysis report (TSAR) in 1996. DOE submitted the TSAR to the United States Nuclear Regulatory Commission (NRC) for review under 10CFR72 and requested that the NRC staff evaluate the TSAR and issue a Safety Evaluation Report (SER) that could be used and referenced by an applicant seeking a site-specific license for the construction and operation of a DTS. DOE also initiated a cold demonstration of major subsystem prototypes in 1996. After careful assessment, the NRC agreed that the DTS concept has merit. However, because the TSAR was not site-specific and was lacking some detailed information required for a complete review, the NRC decided to issue an Assessment Report (AR) rather than a SER. This was issued in November 2000. Additional information that must be included in a future site-specific Safety Analysis Report for the DTS is identified in the AR. The DTS consists of three major sections: a Preparation Area, a Lower Access Area, and a Transfer Confinement Area. The Preparation Area is a sheet metal building where casks are prepared for loading, unloading, or shipment. The Preparation Area adjoins the Lower Access Area and is separated from the Lower Access Area by a large shielded door. The Lower Access Area and Transfer Confinement Area are contained within concrete walls approximately three feet thick. These are the areas where the casks are located and where the fuel is moved during transfer operations. A floor containing two portals separates the Lower Access Area and the Transfer Confinement Area. The casks are located below the floor, and the fuel transfer operation occurs above the floor. The cold demonstration of the DTS was successfully conducted at the Idaho National Engineering and Environmental Laboratory (INEEL) as a cooperative effort between the DOE and EPRI. The cold demonstration was limited to the fuel handling equipment, the cask lid handling equipment, and the cask interface system. The demonstration included recovery operations associated with loss of power or off-normal events. The demonstration did not include cask receiving and lid handling; cask transport and lifting; vacuum/inerting/leak test; canister welding; decontamination; heating, ventilation, and air conditioning; and radiation monitoring. The demonstration test was designed to deliberately challenge the system and determine whether any specific system operation could adversely impact or jeopardize the operation or safety of any other function or system. All known interlocks were challenged. As in all new systems, there were lessons learned during the operation of the system and a few minor modifications made to ease operations. System modifications were subsequently demonstrated. The demonstration showed that the system operated as expected and provided times for normal fuel transfer operations. The demonstration also showed that recovery could be made from off-normal events.

scholarly journals Thermal Aspects of Safety Analysis for Shipment of West Valley Melter

2017 ◽  
Author(s):  
James E. Laurinat ◽  
Matthew R. Kesterson ◽  
Jeffery L. England ◽  
Edward T. Ketusky ◽  
Charles A. McKeel ◽  
...  
Keyword(s):  
National Laboratory ◽  
Safety Analysis ◽  
Department Of Energy ◽  
Savannah River ◽  
The West ◽  
Heating Source ◽  
Savannah River National Laboratory ◽  
Set Up ◽  
High Level ◽  
Cellular Concrete

The thermal aspects of a safety analysis for shipment of the West Valley melter are presented. The West Valley melter was used from 1996 to 2002 to vitrify regionally sourced high level radioactive waste. The U.S. Department of Energy (DOE) set up the West Valley Demonstration Project to encase this melter and grout it in low density cellular concrete, for disposal. DOE-West Valley requested the Savannah River National Laboratory to prepare a Safety Analysis Report. The thermal portion of the safety analysis covers Normal Conditions of Transport (NCT) and Hypothetical Accidents Conditions (HAC), as defined in the Code of Federal Regulations. For NCT, it is assumed that the encased melter is stored in either shade or direct sunlight at an ambient temperature of 311 K (100 °F). The defining HAC is exposure to a 1075 K (1475 °F) fire for 30 minutes. Finite element computer models were used to compute temperature profiles for NCT and HAC, given the thermal properties of the melter and its contents and tabulated radiolytic heating source concentrations. The resulting temperature conditions were used to estimate the pressurization due to evaporation of water from the concrete. The maximum calculated gauge pressures were determined to be 81 kPa (12 psig) for NCT and 580 kPa (84 psig) for HAC.

scholarly journals Examination of a Proposed Validation Data Set Using CFD Calculations

2009 ◽  
Author(s):  
Richard W. Johnson
Keyword(s):  
Cfd Simulation ◽  
Initial Conditions ◽  
Fluid Dynamic ◽  
Safety Analysis ◽  
The United States ◽  
Department Of Energy ◽  
United States Department ◽  
Scaled Model ◽  
Validation Data ◽  
Data Set

The United States Department of Energy is promoting the resurgence of nuclear power in the U. S. for both electrical power generation and production of process heat required for industrial processes such as the manufacture of hydrogen for use as a fuel in automobiles. The DOE project is called the next generation nuclear plant (NGNP) and is based on a Generation IV reactor concept called the very high temperature reactor (VHTR), which will use helium as the coolant at temperatures ranging from 450 °C to perhaps 1000 °C. While computational fluid dynamics (CFD) has not been used for past safety analysis for nuclear reactors in the U. S., it is being considered for safety analysis for existing and future reactors. It is fully recognized that CFD simulation codes will have to be validated for flow physics reasonably close to actual fluid dynamic conditions expected in normal and accident operational situations. To this end, experimental data have been obtained in a scaled model of a narrow slice of the lower plenum of a prismatic VHTR. The present article presents new results of CFD examinations of these data to explore potential issues with the geometry, the initial conditions, the flow dynamics and the data needed to fully specify the inlet and boundary conditions; results for several turbulence models are examined. Issues are addressed and recommendations about the data are made.

Implementation of the ASME Code and NQA-1 in U.S. Department of Energy Packaging Certification Program Training Courses

2011 ◽  
Author(s):  
Lawrence F. Gelder
Keyword(s):  
Safety Analysis ◽  
Department Of Energy ◽  
Department Of Transportation ◽  
Type B ◽  
Certification Program ◽  
Radioactive Materials ◽  
Package Design ◽  
Safety Standards ◽  
Asme Code ◽  
Training Courses

Under the authorization of the Department of Transportation, per 49 CFR Part 173.7(d), Type B and fissile radioactive materials packagings made by or under the direction of the U.S. Department of Energy (DOE) may be used for the transportation of Class 7 materials when evaluated, approved, and certified by DOE against packaging standards equivalent to those specified in 10 CFR Part 71. The DOE certificate is issued on the basis of a safety analysis report of the package design and application. The applicant must demonstrate to DOE the package meets the standards in the 10 CFR Part 71. Since the Type B and fissile radioactive materials packaging standards specified in 10 CFR Part 71 are performance based standards, guides and other tools are necessary to demonstrate how a package design meets the standards. Two essential tools used by packaging applicants and reviewers to quantify and demonstrate compliance with the safety standards/requirements of the CFR are the ASME Boiler and Pressure Vessel (B&PV) Code and ASME NQA-1. The DOE Packaging Certification Program develops and sponsors training courses for packaging applicants and reviewers. Many of these courses are required training by DOE for persons that manage or prepare safety analysis reports for package designs (i.e., applications) submitted to the DOE for certification. The ASME B&PV Code and NQA-1 are ubiquitous in the DOE core training courses. This paper provides an overview how the ASME B&PV Code and NQA-1 are implemented in DOE Packaging Certification Program training courses.

Safety Basis Design Development Challenges

2007 ◽  
Author(s):  
Jonathan Young ◽  
Pete Lowry ◽  
Bruce Schmitt ◽  
Robin Sullivan ◽  
Lenna Mahoney ◽  
...  
Keyword(s):  
Spent Nuclear Fuel ◽  
Safety Management ◽  
Safety Analysis ◽  
Nuclear Safety ◽  
Department Of Energy ◽  
Design Activity ◽  
Design Development ◽  
Design Basis ◽  
Safety Features ◽  
System 1

“Designing in Safety” is a desired part of the development of any new potentially hazardous system, process, or facility. It is a required part of nuclear safety activities as specified in the U.S. Department of Energy (DOE) Order 420.1B, Facility Safety. This order addresses the design of nuclear related facilities developed under federal regulation 10CFR830, Nuclear Safety Management. 10CFR830 requires that safety basis documentation be provided to identify how nuclear safety is being adequately addressed as a condition for system operation (e.g., the safety basis). To support the development of the safety basis, a safety analysis is performed. Although the concept of developing a design that addresses safety is simple, the execution can be complex and challenging. This paper addresses those complexities and challenges for the design activity of a system to treat sludge, a corrosion product of spent nuclear fuel, at DOE’s Hanford Site in Washington State. The system being developed is referred to as the Sludge Treatment Project (STP). This paper describes the portion of the safety analysis that addresses the selection of design basis events using the experience gained from the STP and the development of design requirements for safety features associated with those events. Specifically, the paper describes the safety design process and the application of the process for two types of potential design basis accidents associated with the operation of the system, 1) flashing spray leaks and 2) splash and splatter leaks. Also presented are the technical challenges that are being addressed to develop effective safety features to deal with these design basis accidents.

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