Launch Safety Analysis for Radioisotope Power Systems.
Abstract not provided.
NESDPS Office of Nuclear Energy Space and Defense ↗
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Databook Inputs to Safety Analysis for Launch Approval
The purpose of this document is to define the data inputs required for a full safety analysis to support launch approval.
97 MATHEMATICS AND COMPUTING↗
Modeling Impact-Induced Reactivity Changes Using DAG-MCNP5
There is a long literature studying the criticality of space reactors immersed in water/sand after a launch accident; however, most of these studies evaluate nominal or uniformly compacted system configurations. There is less research on the reactivity consequences of impact, which can cause large structural deformation of reactor components that can result in changes in the reactivity of the system. Predicting these changes is an important component of launch safety analysis. This paper describes new features added to the DAG-MCNP5 neutronics code that allow the criticality analysis of deformed geometries. A CAD-based solid model of the reactor geometry is used to generate an initial mesh for a structural mechanics impact calculation using the PRONTO3D/PRESTO continuum mechanics codes. Boundary conditions and material specifications for the reactivity analysis are attached to the solid model that is then associated with the initial mesh representation. This geometry is then updated with the deformed finite element mesh to perturb node coordinates. DAG-MCNP5 was extended to accommodate two consequences of the large structural deformations: dead elements representing fracture, and small overlaps between adjacent volumes. The dead elements are removed during geometry initialization and adjustments are made to conserve mass. More challenging, small overlaps where adjacent mesh elements contact cause the geometric queries to become unreliable. A new point membership test was developed that is tolerant of self-intersecting volumes, and the particle tracking algorithm was adjusted to enable transport through small overlaps. These new features enable DAG-MCNP5 to perform particle transport and criticality eigenvalue calculations on both deformed mesh geometry and CAD geometry with small geometric defects. Detailed impact simulations were performed on an 85-pin space reactor model. In the most realistic model that included NaK coolant and water in the impact simulation, the eigenvalue was determined to increase 2.7% due to impact.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Consequence Analysis for Mars Science Laboratory Launch Safety Study.
Abstract Not Provided
NESDPS Office of Nuclear Energy Space and Defense ↗
Consequence Analysis for Mars Science Laboratory Launch Safety Study.
Abstract not provided.
NESDPS Office of Nuclear Energy Space and Defense ↗
SFM - Sandia Fireball Model
SAND2022-13231 O Sandia Fireball Model is used in the assessment of liquid propellant fireballs in the source term analysis for space nuclear systems launch safety. Sandia National Laboratories is a multimission laboratory managed and operated by National Technology & Engineering Solutions of Sandia, LLC, a wholly owned subsidiary of Honeywell International Inc., for the U.S. Department of Energy’s National Nuclear Security Administration under contract DE-NA0003525.
Powers, Dana↗
New Design for an HLW Repository (for Spent Fuel and Waste from Reprocessing) in a Salt Formation in Germany - 12213
In autumn 2010, after a 10-year moratorium, exploration was resumed in Gorleben, the potential site for a German HLW repository. At the same time, the Federal Government launched a two-year preliminary safety analysis to assess whether the salt dome at Gorleben is suitable to host all heat-generating radioactive waste generated by German NPPs based on the waste amounts expected at that time. The revised Atomic Energy Act of June 2011 now stipulates a gradual phase-out of nuclear energy production by 2022, which is 13 years earlier than expected in 2010. A repository design was developed which took into account an updated set of data on the amounts and types of expected heat-generating waste, the documented results of the exploration of the Gorleben salt dome, and the new 'Safety Requirements Governing the Final Disposal of Heat-Generating Radioactive Waste' of 30 September, 2010. The latter has a strong influence on the conceptual designs as it requires that retrievability of all waste containers is possible within the repository lifetime. One design considered that all waste containers will be disposed of in horizontal drifts of a geologic repository, while the other design considered that all waste containers will be disposed of in deep vertical boreholes. For both options (emplacement in drifts/emplacement in vertical boreholes), the respective design includes a selection of waste containers, the layout of drifts, respectively lined boreholes, a description of emplacement fields, and backfilling and sealing measures. The design results were described and displayed and the differences between the two main concepts were elaborated and discussed. For the first time in both repository designs the requirement was implemented to retrieve waste canisters during the operational phase. The measures to fulfill this requirement and eventually the consequences were highlighted. It was pointed out that there arises the need to keep transport- and storage casks in adequate numbers and interim storage facilities available until the repository is closed. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Health Physics Parameter Analysis for the Mars 2020 Launch
This report describes research into three health physics parameters used by Launch Safety (LS) for which the appropriate value, distribution, or applicability came into question during preparation of the Mars 2020 LS analysis. These parameters and associated issues include the Dose and Dose Rate Effectiveness Factor (DDREF) and its use in health effects calculations, a methodology for translating projected contamination per unit area into dose to aquatic and terrestrial biota, and plutonium transfer factors for use in ingestion pathway consequence analyses.
61 RADIATION PROTECTION AND DOSIMETRY↗
Benchmark Modeling and Simulation of the FFTF LOFWOS Test #13 Using SAM
The Fast Flux Test Facility (FFTF) was a 400 MW thermal powered, oxide-fueled, liquid sodium cooled test reactor, built to assist development and testing of advanced fuels and materials for fast breeder reactors. In July 1986, a series of unprotected Loss of Flow Without Scram (LOFWOS) transients were performed in FFTF as part of the Passive Safety Testing (PST) program. The LOFWOS Test #13, which was initiated at 50% power and 100% flow with the pump pony motors left off, has been chosen as a benchmark case by IAEA to support collaborative efforts within international partnerships on the validation of simulation tools and models in the area of sodium fast reactor passive safety in an IAEA Coordinated Research Project (CRP), launched in October 2018. The System Analysis Module (SAM) is an advanced and modern system analysis tool under development at Argonne National Laboratory for advanced non-LWR safety analysis. It utilizes the object-oriented application framework MOOSE to leverage the modern software environment and advanced numerical methods. The capabilities of SAM are being extended to enable the transient modeling, analysis, and design of various advanced nuclear reactor systems. To participate the IAEA CRP and enhance the SAM validation base for advanced reactor transient safety analysis, benchmark simulations of the FFTF LOFWOS Test #13 are performed using the SAM code. In this first phase of the validation effort, the thermal-hydraulic behavior of the reactor system is the focus and the reactor kinetics is not considered in the SAM FFTF model. Instead, the results of Argonne’s neutronics calculations are directly used, including the power shape of the active core region and the power history during the transient. The simulation results of FFTF at steady state agreed well with the measured data from the test. During the transient, reasonably good agreement were also obtained. Future work to improve the model will focus on introducing the reactivity predictions into the model, as well as better understanding or resolving the current discrepancies with the measured data.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Comparison of Light-Weight Radioisotope Heater Unit (LWRHU) Historical Testing to Special Form Testing Requirements
This paper compares the Light-Weight Radioisotope Heater Unit (LWRHU) historical testing to Department of Transportation (DOT) Special Form criteria per 49 CFR Part 173, Shippers-General Requirements for Shipments and Packages, Section 469, Tests for special form Class 7 (radioactive) materials as specified in DOE-STD-1027, Hazard Categorization of DOE Nuclear Facilities. Compliance with the aforementioned criteria allows for the exclusion of radioactive sources from the facility’s nuclear material inventory. As part of a new program at Los Alamos National Laboratory (LANL), LWRHUs are planned to be moved into TA-55, PF-5 Power Source Assembly Area (PSAA) for assembly and testing activities. The PSAA Facility is categorized as a low hazard, non-nuclear facility per DOE STD-1027 and is described in TA55-PF-5-FHC-11-001, Facility Hazard Categorization of the PSAA at PF-5, TA-55. The Facility Hazard Categorization document states, there must be documentation that radioactive sources are engineered to pass the special form testing specified by DOT in 49 CFR 173.469, Tests for special form Class 7 (radioactive) materials or ANSI N43.6, Sealed Radioactive Sources - Classification. This paper documents the extensive historical environmental safety testing of the LWRHU and demonstrates its ability to survive potential launch safety accident scenarios. The testing results are used herein to demonstrate through a comparative analysis that the robust design of the LWRHU meets or exceeds the special form testing specifications in 49 CFR 173.469, enabling the LWRHU to be transported and handled in PF-5 in support of testing and assembly program.
07 ISOTOPE AND RADIATION SOURCES↗
Cost Comparison in 2015 Dollars for Radioisotope Power Systems -- Cassini and Mars Science Laboratory
Radioisotope power systems (RPSs) have enabled missions requiring reliable, long-lasting power in remote, harsh environments such as space since the early 1960s. Costs for RPSs are high, but are often misrepresented due to the complexity of space missions and inconsistent charging practices among the many and changing participant organizations over the years. This paper examines historical documentation associated with two past successful flight missions, each with a different RPS design, to provide a realistic cost basis for RPS production and deployment. The missions and their respective RPSs are Cassini, launched in 1997, that uses the general purpose heat source (GPHS) radioisotope thermoelectric generator (RTG), and Mars Science Laboratory (MSL), launched in 2011, that uses the multi-mission RTG (MMRTG). Actual costs in their respective years are discussed for each of the two RTG designs and the missions they enabled, and then present day values to 2015 are computed to compare the costs. Costs for this analysis were categorized into two areas: development of the specific RTG technology, and production and deployment of an RTG. This latter category includes material costs for the flight components (including Pu-238 and fine weave pierced fabric (FWPF)); manufacturing of flight components; assembly, testing, and transport of the flight RTG(s); ground operations involving the RTG(s) through launch; nuclear safety analyses for the launch and for the facilities housing the RTG(s) during all phases of ground operations; DOE’s support for NEPA analyses; and radiological contingency planning. This analysis results in a fairly similar 2015 normalized cost for the production and deployment of an RTG—approximately $118M for the GPHS-RTG and $109M for the MMRTG. In addition to these two successful flight missions, the costs for development of the MMRTG are included to serve as a future reference. Note that development costs included herein for the MMRTG do not include costs from NASA staff or facilities for their development efforts—they only include the amounts costed by DOE and DOE contractors. The 2015 value for MMRTG development is $83M. Both of the RPS types analyzed herein use the general purpose heat source (GPHS) module as the “heart of the RPS.” The estimates presented herein do not include development costs for the GPHS. These estimates also do not include the RPS infrastructure cost to maintain the facilities, equipment, and personnel necessary to enable the production of RPSs, except to the extent that the infrastructure is utilized during the production campaigns to provide RPSs for missions. It was not until after the Cassini mission that an RPS infrastructure funding structure was defined and funded separately from mission-specific elements. The information presented herein could allow for more accurate budget planning estimates for space missions being considered over the next decade and beyond.
07 ISOTOPE AND RADIATION SOURCES↗
Kilopower Space Reactor Launch Safety - Maximum Credible Dose for a Criticality Accident
This note examines potential consequences for the Kilopower small space reactor going critical during a launch accident. The goal of this analysis is to postulate the dose at a distance that could be generated during a criticality accident event (with emphasis on potential public doses.) This note does not examine local consequences to an individual who would be close enough to the reactor (within several 10’s of meters) to receive a lethal dose from the direct shine of neutrons and gamma radiation present when the reactor is critical.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Kilopower Space Reactor Launch Safety - Maximum Credible Fissions for a Criticality Accident
This note examines potential accident conditions for the Kilopower space reactor going critical during a launch accident. The goal of this analysis is to postulate the total number of fissions that could be generated during an accident event. The total number of fissions is an approximate surrogate for what will be the maximum release (or source term) for the accident. From the source term an approximation of dose (based upon the fission products generated and released) can be estimated. This note will only attempt to app roximate the total number of fissions to the right order of magnitude. Dose consequences from the number of fissions will be documented in a separate note.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Collaborations of the NEAMS Center of Excellence in 2025
The mission of the NEAMS Center of Excellence for Thermal-fluids Applications in Nuclear Energy is to advance the goals of the NEAMS program by providing leadership, best practices, research, and support and training for computational thermal hydraulics. In particular NEAMS tools are being developed to ultimately span the entire range of length- and time-scales required for a comprehensive reactor design and safety analysis. The focus of The Center is on the use, best practices, and deployment to stakeholders of the thermal hydraulic codes. The Center of Excellence, originally launched in 2018, has as its key goal to serve as a “front door” to industry. As part of this goal, The Center is in its sixth year of a program to engage in collaborative efforts between the laboratories and industry with the objective of stimulating cooperation and increasing the adoption of thermal hydraulics tools developed under NEAMS by the industry at large. With an overwhelmingly positive response from participants, previous short-term projects have led to more in-depth collaborations.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Sampling and Analysis Plan for Investigation of Undeclared Nuclear Activities - 20104
The nuclear-centered energy policy in South Korea is moving towards a nuclear-free era. It also facilitates an action or progress on the denuclearization of the Korean peninsula. The Nuclear Safety and Security Commission (NSSC) launched a new R and D project which attempts to verify nuclear activities via analyzing environmental samples. Environmental samples possibly contain significant radiological information relevant to past undeclared activities. In order to precisely verify them, it is essential to provide 1) the concrete information of target facilities, 2) appropriate methodologies and procedures, and 3) well-suited instrumentation and equipment corresponding to sample signatures. It is also important to provide scientific reliability or validity in any process such as sample collection, handling, transportation and laboratory analysis. Therefore, the information of nuclear facilities (fuel fabrication plant, graphite moderated reactor, reprocessing plant) in North Korea are reviewed in order to determine parameters for appropriate sampling methods representing a facility's characteristics. Secondly, a sampling procedure for radioactive elements is being developed based on international and domestic standards such as ASTM (American Society for Testing and Materials), KS (Korean industrial Standards), and International Atomic Energy Agency (IAEA) manuals. The procedure contains some specific information including sampling process, sampling tools, amount of sample, sample identification, and other related information. Developing the qualified national guidelines is most important since sampling and analyzing results can be used as evidences of nuclear activities. The guideline developed through this work will clearly define various methodologies such as document verification, sampling procedures, collection methods, sample storage criteria, chain of custody, and transportation conditions. The final objective of this work is to establish the infrastructure and regulatory framework for assessment of nuclear activities, and provide a national guideline for sampling and analysis. This will bring confidence in the results and play a critical role in complete denuclearization progress in South Korea. (authors)
98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
Mars 2020 Risk Integration and Uncertainty Working Group Status Report, FY 2017
This report provides a list of accomplishments of the Mars 2020 Risk Integration and Uncertainty Working Group for the fiscal year 2017 which includes participation in the Technical Information Meeting #2, review and comments on the methodologies and results provided in the Mars 2020 Safety Analysis Report (SAR) Draft Databook, development of a Probabilistic Risk Assessment model in the risk assessment program SAPHIRE using the data provided in the SAR Databook, and an independent launch risk analysis based on the Mars 2020 launch vehicle’s lineage and design similarity to other launch vehicles.
99 GENERAL AND MISCELLANEOUS↗
Decontamination and Decommissioning Activities of a Radiopharmaceuticals Production Facility - 18067
On May 2012, Best Medical Belgium S.A. (BMB), a manufacturer of pharmaceutical products located on the nuclear site of Fleurus, was declared bankrupt. According to the Belgian law, ONDRAF/NIRAS (Belgian Radioactive Waste Management Agency) was entrusted with the clean-up and decommissioning operations of the former BMB facilities, nowadays known as 'ONDRAF Site Fleurus' (ONSF). In June 2015, ONSF launched a Decontamination and Decommissioning (D and D) program, for which Tractebel: - Supports the definition of the clean-up and decommissioning strategies; - Drafts the Final Decommissioning Plan (FDP) and Safety Analysis Report (SAR) that will be submitted to FANC/AFCN (Belgian Federal Agency for Nuclear Control) in order to get the License for decommissioning; - Drafts the technical specifications of the public tendering process related to the D and D program; - Ensures the day to-day follow-up of the D and D activities on site. Four nuclear buildings are part of the D and D activities. One of them, building 14 (B14), represents the most challenging and extensive task with the clean-up of highly contaminated areas and its further decommissioning up to the final state. This building features two cyclotrons (CGR and IBA cyclotrons) successively used to produce radioisotopes, as well as hot cells and glove boxes used for their purification and the conditioning of the final pharmaceutical product. The present paper aims at presenting the most recent development steps of the clean-up phase regarding the adaptations of existing facilities and the upcoming activities in preparation of the decommissioning activities. To allow the sorting and the treatment of the radioactive material identified in the physical and radiological inventory, several workshops have been implemented within the B14 building in order to deal with radioactive materials arising from the clean-up and D and D activities. Following the conceptual design phase, Tractebel has actively monitored the on-site activities for the construction of those workshops such as: - The modification of the civil works; - The revamping of the ventilation system; - The installation of dedicated lifting equipment and workstations equipped with cutting devices; - The assembly of a docking station for filling the radioactive packages. The removal of the activated concrete surrounding the IBA cyclotron requires the erection of a containment structure and an extension structure. The containment structure, composed of light metallic beams, is located above the cyclotron and its irradiation zones. For operational security reasons and working efficiency, the cutting process of the concrete ceilings and walls will be performed from the top inside the containment structure by means of an eight tons overhead bridge crane. Concrete blocks resulting from the cutting process will then be lifted, without any loss of containment, towards the extension zone of the nuclear zone of the IBA cyclotron for further resizing, sorting and evacuation towards the identified waste routes. To prepare this task which will be performed during the decommissioning phase, Tractebel sets up the necessary technical specifications. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗