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At least 757 records · Page 42

Steady state RANS simulations of temperature fluctuations in single phase turbulent mixing

Single phase turbulent mixing in nuclear power plant circuits where a strong temperature gradient is present is known to precipitate pipe failure due to thermal fatigue. Experiments in a square mixing channel offer the opportunity to study the phenomenon under simple and easily reproducible boundary conditions. Measurements of this kind have been performed extensively at the Paul Scherrer Inst. in Switzerland with a high density of instrumentation in the Generic Mixing Experiment (GEMIX). As a fundamental mixing phenomena study closely related to the thermal fatigue problem, the experimental results from GEMIX are valuable for the validation of CFD codes striving to accurately simulate both the temperature and velocity fields in single phase turbulent mixing. In the experiments two iso-kinetic streams meet at a shallow angle of 3 degrees and mix in a straight channel of square cross-section under various degrees of density, temperature, and viscosity stratification over a range of Reynolds numbers ranging from 5*10{sup 3} to 1*10{sup 5}. Conductivity measurements, using wire-mesh and wall sensors, as well as optical measurements, using particle image velocimetry, were conducted with high temporal and spatial resolutions (up to 2.5 kHz and 1 mm in the case of the wire mesh sensor) in the mixing zone, downstream of a splitter plate. The present paper communicates the results of RANS modeling of selected GEMIX tests. Steady-state CFD calculations using a RANS turbulence model represent an inexpensive method for analyzing large and complex components in commercial nuclear reactors, such as the downcomer and reactor pressure vessel heads. Crucial to real world applicability, however, is the ability to model turbulent heat fluctuations in the flow; the Turbulent Heat Flux Transport model developed by ANSYS CFX is capable, by implementation of a transport equation for turbulent heat fluxes, of readily modeling these values. Furthermore, the closure of the turbulent heat flux transport equation evokes a transport equation for the variance of the enthalpy. It is therefore possible to compare the modeled fluctuations of the liquid temperature directly with the scalar fluctuations recorded experimentally with the wire-mesh. Combined with a working Turbulent Heat Flux Transport model, complex mixing problems in large geometries could be better understood. We aim for the validation of Reynolds Stress based RANS simulations extended by the Turbulent Heat Flux Transport model by modeling the GEMIX experiments in detail. Numerical modeling has been performed using both BSL and SSG Reynolds Stress Models in a test matrix comprising experimental trials at the GEMIX facility. We expand on the turbulent mixing RANS CFD results of (Manera 2009) in a few ways. In the GEMIX facility we introduce density stratification in the flow while removing the characteristic large scale vorticity encountered in T-junctions and therefore find better conditions to check the diffusive conditions in the model. Furthermore, we study the performance of the model in a very different, simpler scalar fluctuation spectrum. The paper discusses the performance of the model regarding the dissipation of the turbulent kinetic energy and dissipation of the enthalpy variance. A novel element is the analyses of cases with density stratification. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Evaluation of effective-stress-function algorithm for nuclear fuel simulation

In a pressurized water reactor (PWR), the mechanical integrity of nuclear fuel is the most critical issue as it is an important barrier for fission products released into the environment. The integrity of zirconium cladding that surrounds uranium oxide can be threatened during off-normal operation owing to a pellet-cladding mechanical interaction (PCMI). To analyze the fuel and cladding behavior during off-operation, the fuel performance code should calculate an inelastic analysis in two - or three-dimensional calculations. In this paper, the effective stress function (ESF) algorithm based on a two-dimensional FE module has been implemented to simulate the inelastic behavior of the cladding with stability and accuracy. The ESF algorithm solves the governing equations of the inelastic constitutive behavior by calculating the zero of the appropriate effective-stress-function. To verify the accuracy of the ESF algorithm for an inelastic analysis, a code-to-code benchmark was performed using the commercial FE code, ANSYS 13.0. To demonstrate the stability and convergence of the implemented algorithm, the number of iterations in the ESF algorithm was compared with that in a sequential algorithm in the case of an inelastic problem. Consequently, the evaluation results demonstrate that the implemented ESF algorithm improves the efficiency of the computation without a loss of accuracy for an inelastic analysis. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Mixed-oxide fuel decay heat analysis for BWR LOCA safety evaluation

The mixed-oxide (MOX) fuel decay heat behavior is analyzed for Boiling Water Reactor (BWR) Loss of Coolant Accident (LOCA) safety evaluation. The physical reasoning on why the decay heat power fractions of MOX fuel fission product (FP) are significantly lower than the corresponding decay heat power fractions of uranium-oxide (UOX) fuel FP is illustrated. This is primarily due to the following physical phenomena. -The recoverable energies per fission of plutonium (Pu)-239 and Pu-241 are significantly higher than those of uranium (U)-235 and U-238. Consequently, the fission rate required to produce the same amount of power in MOX fuel is significantly lower than that in UOX fuel, which leads to lower subsequent FP generation rate and associated decay heat power in MOX fuel than those in UOX fuel. - The effective FP decay energy per fission of Pu-239 is significantly lower than the corresponding effective FP decay energy per fission of U-235, e.g., Pu-239's 10.63 Mega-electron-Volt (MeV) vs. U-235's 12.81 MeV at the cooling time 0.2 second. This also leads to lower decay heat power in MOX fuel than that in UOX fuel. The FP decay heat is shown to account for more than 90% of the total decay heat immediately after shutdown. The FP decay heat results based on the American National Standard Institute (ANSI)/American Nuclear Society (ANS)-5.1-1979 standard method are shown very close to the corresponding FP decay heat results based on the ANSI/ANS-5.1-2005 standard method. The FP decay heat results based on the ANSI/ANS-5.1-1979 simplified method are shown very close to but mostly slightly lower than the corresponding FP decay heat results based on the ANSI/ANS-5.1-1971 method. The FP decay heat results based on the ANSI/ANS-5.1-1979 simplified method or the ANSI/ANS-5.1-1971 method are shown significantly larger than the corresponding FP decay heat results based on the ANSI/ANS-5.1-1979 standard method or the ANSI/ANS-5.1-2005 standard method. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

The MacArthur Maze Fire and Roadway Collapse: Consequences for SNF Transportation - 12476

In 2007, a severe transportation accident occurred near Oakland, California, on a section of Interstate 880 known as the 'MacArthur Maze', involving a tractor trailer carrying gasoline which impacted an overpass support column and burst into flames. The subsequent fire caused the collapse of portions of the Interstate 580 overpass onto the remains of the tractor-trailer in less than 20 minutes, due to a reduction of strength in the structural steel exposed to the fire. The US Nuclear Regulatory Commission is in the process of examining the impacts of this accident on the performance of a spent nuclear fuel transportation package, using detailed analysis models, in order to determine the potential regulatory implications related to the safe transport of spent nuclear fuel in the United States. This paper will provide a summary of this effort and present results and conclusions. The detailed thermal models of the MacArthur Maze fire scenario with ANSYS and COBRASFS have produced preliminary results indicating that in a fire of this severity, the peak fuel cladding temperature would almost certainly exceed the short-term limit of 570 deg. C (1058 deg. F), and would likely exceed the Zircaloy burst temperature limit of 750 deg. C (1382 deg. F) assumed in previous transportation studies. Additional work is needed to refine and verify some of the details of these complex models, but the overall results are consistent with previous fire analyses with similar models, and with the results obtained for the HAC fire evaluations with these models. These results as well as future results produced by these models can therefore be considered as reliable estimates of the temperatures that would be experienced in fire conditions of the severity of the MacArthur Maze fire scenario. The structural analyses show that the GA-4 package is robust enough to withstand the impact of the overhead span without suffering major damage or deformation to the containment boundary. The greatest potential for local package damage in this scenario appears to be at the bolted closure end. The thermal expansion response of the closure bolts and impact limiter attachment bolts were evaluated in a separate analysis, and it was determined that both the lid and impact limiters are expected to remain in place. The response of the closure seal is currently being evaluated separately in the context of the accident's overall potential to release radioactive material in the environment. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Performance of an RPM based on Gd-lined plastic scintillator for neutron and gamma detection [ANIMMA--2015-IO-372]

A Radiation Portal Monitor (RPM) was developed by the Istituto Nazionale di Fisica Nucleare (INFN) and Ansaldo Nucleare (ANN) within the FP7 SCINTILLA European project. The system was designed to detect both gamma and neutron radiation with a single technology. It is conceived to monitor vehicle and cargo containers in transits across borders or ports, to find radioactive elements and to avoid illegal trafficking of strategic nuclear materials. The system is based on a {sup 3}He-free neutron detection technology using plastic scintillators coupled to Gadolinium to detect and discriminate gamma from neutron signals. During the 3 years of the SCINTILLA project the construction and test of the first two prototypes drove the definition of the final layout of a full RPM system consisting of two twin pillars as a portal for vehicle and cargo container scan. A custom System Control Software (SCS) manages the electronics of the RPM, the ancillary devices and the data analysis. The combination of the detector layout and of the software functionalities enables both to distinguish neutrons and gammas and to identify the energy range of a detected gamma source. The system was initially characterized via static tests with gamma and neutron sources in the INFN laboratory. These measurements were used to calibrate the detector, evaluate the response of the single pillars as well as of the full system, and optimize the RPM configuration and discrimination algorithm. During this phase, specific tests were performed to study the stability over time of the system, monitoring the measured the neutron and gamma count rates over periods of several weeks. The results allow us to demonstrate the reliability and robustness of the RPM. In a second time the RPM performance was studied via dynamic tests performed during the SCINTILLA test and benchmark campaigns. These measurements took place in the JRC ITRAP+10 facility at Ispra (Varese-Italy). The laboratory is equipped with an experimental set-up for dynamic tests of multiple systems according to international standards. The performed measurements utilized radioactive sources with activities selected according to ANSI and IEC standards to test the detector alarm performances in terms of gamma and neutron response, sensitivity to high gamma fields, sensitivity to moderated neutron sources as well as false alarm rates (FAR). In addition, the RPM was tested in challenging configurations exceeding the requirements set by international standards to determine the real limits of the system. The results obtained during these campaigns demonstrated that the system detection efficiency is not only compliant to international standards for its category, but often exceeds them, demonstrating the validity of the chosen technology and of the implemented layout. The positive performance also showed the effectiveness of the SCS and of its functionalities. To further demonstrate the system capabilities, a test in a real-life environment of the RPM is planned to happen in a near future by installing the detectors in a seaport. In this presentation I will give an overview of the RPM characteristics, of its performances as determined in the test campaign mentioned above and of future plans, to demonstrate how this technology can be an effective choice for the realization of {sup 3}He-free RPM detectors. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Hydraulic performance evaluation of rod bundle with a concept spacer grid design

In a pressurized water reactor the spacer grid has a great influence on the thermal-hydraulic and mechanical properties of the fuel assembly. In order to improve the overall performance of the fuel assembly, a new concept spacer grid design names CATF with 'non-bending' guide vane is proposed. The hydraulic performance of the CATF grid design is evaluated by using a commercially computational fluid dynamics (CFD) code, ANSYS CFX12.1, and the evaluation result is compared with ordinary spacer grid including bending guide vane. Three dimensional rod bundles including CATF spacer grid and ordinary spacer grid have been simulated. The analysis results show that local resistance coefficient of CATF spacer grid is smaller than that of ordinary spacer grid by about 29.7% because of the non-bending guide vanes. Furthermore, the cross flow strength between subchannels near the outer strap area, the swirl strength and cross flow strength in subchannels near the outer strap area in 1-10 times hydraulic diameter downstream the spacer grid area of the ordinary grid, are larger than that of CATF spacer grid. The two types of spacer grids appear to have the same mixing characteristics in inner-grid area. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Screening sensitivity analysis of a PWR fuel assembly FEM structural model

Within the present project, computational methods to describe the static mechanical response of PWR fuel assemblies (FA) in the reactor core are developed. For this purpose a structural model of a PWR fuel assembly has been implemented with the Finite Element Method (FEM) using the ANSYS APDL Mechanical code. The purpose of the model is to investigate the long-term creep deformation of the fuel assemblies in the reactor core, commonly known as fuel assembly bow. FA bow is considered to be a complex process with a large number of influencing mechanisms and many only roughly known boundary conditions. Uncertainty and sensitivity analyses are a common way to investigate the predictive power and reliability of such complex models. Hence, as a first approach to understand the importance of the many model parameters, a qualitative sensitivity analysis of the model was conducted using the screening method proposed by Morris. In the first part of the paper, the FA structural model and its principal parameters are presented. The second part presents first the different simulation experiments used for the screening analyses. Based on the model description, relevant input parameters are then identified and a typical range of values is determined. The sensitivity analysis is performed in two subsequent steps. First, the model is screened for influential stiffness parameters with the objective of stiffening the FA structure. Second, the influence of the uncertain boundary conditions and material models on the FA creep deformation is tested. For the considered FA design it is concluded that the influence of the involved uncertainties is comparable to the influence of the parameters which can increase the stiffness. An integral approach accounting for these uncertainties is hence necessary to predict FA bow reliably in order to take additional measures for a further reduction of FA bow amplitudes. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Terminal Sinking Velocity for Waste Packages Falling in a Deep Borehole - 17063

This evaluation of terminal velocity for a waste package sinking in a deep borehole filled with emplacement fluid was done to support concept development for the Deep Borehole Field Test (DBFT). The DBFT objective is to evaluate the safety and feasibility of the deep borehole disposal (DBD) concept for certain radioactive wastes. The DBD concept calls for emplacing waste packages in the lower 2 km of a borehole drilled into crystalline basement rock to a depth of 5 km. Previous studies identified that accidentally dropping one or more waste packages during emplacement could be an important risk, particularly if a waste package is breached. This study evaluated the range of terminal sinking velocity, which can be used to design waste packages and impact limiters to ensure that breach does not occur. Based on an analytical model of a sinking container, a highly turbulent flow is expected to develop in the emplacement fluid. Turbulence is a complex three-dimensional phenomenon and the applicability of analytical models is limited. To backstop the analytical model and to extend predictions to other package dimensions and fluid properties, numerical simulation was performed using the ANSYS Fluent computational fluid dynamics code. Terminal velocity was calculated for downhole temperatures up to 1200 C, and three emplacement fluids (water, NaCl brine, and NaBr brine). Simulation results confirmed that flow around the sinking container is highly turbulent (Reynolds number 5.6 x 10{sup 4} to 7.3 x 10{sup 5}). Turbulence increases with temperature and decreases with density. Calculated terminal velocity ranges from 1.95 (200 C) to 2.13 m/s (1200 C) in water; 1.61 (200 C) to 1.79 m/s (1200 C) in NaCl; and 1.30 (200 C) to 1.46 m/s (1200 C) in NaBr. Terminal velocity is inversely related to fluid density, and the temperature effect on density appears to have a stronger effect on terminal velocity than the temperature effect on viscosity. The foregoing results are for non-perforated casing, but in practice the casing may be perforated to prepare for cementing, and to allow relief of fluid pressure from thermal expansion after emplacement of heat generating waste, and after final borehole closure. To represent the effect from perforations, the analytical solution was modified to include bypass flow through perforations and in the annulus. A workable perforation scheme is identified, consistent with borehole construction and closure requirements, using 2-cm and 5-cm diameter perforations. This scheme is estimated to limit the terminal velocity to 3 m/s, for reference waste packages falling in water at 400 C. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Development of a New 48 Inch UF6 Cylinder Overpack - 17476

The Kevil Kougar overpack is new generation USA DOT Specification 7A packaging designed to transport ANSI N14.1 and ISO-7195 compliant 48-inch UF6 cylinders. In December 2016, the USA DOT issued USA/0815/H(U)-96 for the transport of 48X and 48Y cylinders and USA/0819/H(M)-96 for the transport of 48G, 48H, 48HX, 48O, 48OM, 48OM(Allied) and 48T cylinders. The H(U) certificate is for transport of thick-walled cylinders and the H(M) certificate is for transport of thin-walled cylinders. The Kevil Kougar design provides thermal and impact protection for the cylinder under conditions required for a Type A package capable of transporting UF{sub 6} via road, rail, and ocean. The packaging design is a smooth octagonal shape designed to carry one 48-inch UF{sub 6} cylinder. It consists of an upper subassembly and lower subassembly, which are fastened together with ten toggle bolts, five per side. Each subassembly consists of an outer and inner steel shell secured to a frame assembly, with wood and polyurethane foam filling the space between. The protruding components are only for ease of operation and consist of four lifting lugs and fork pockets on the upper body, and the base and fork pockets on the lower body. The design incorporates interior rubber pads for cylinder cushioning, support and alignment, steel stabilizing beams fore and aft to prevent cylinder movement, closure mechanisms, a tamper indicating device and tie-down features. The Kevil Kougar is currently designed to transport contents of depleted uranium and natural uranium. Future activities include licensing the package as Type AF to facilitate transport of historic USA Department of Energy fissile material packages. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

The Publication and Endorsement of the ASME/ANS Probabilistic Risk Assessment Standard for Advanced Non-Light Water Reactor Nuclear Power Plants

Due to growing interest in non-light water reactor (NLWR) concepts, the predecessor to the American Nuclear Society (ANS)/American Society of Mechanical Engineers (ASME) Joint Committee on Nuclear Risk Management (JCNRM) initiated development of a NLWR probabilistic risk assessment (PRA) standard in 2006. The developed standard, released for trial use in 2013, was the first integral PRA standard created by the JCNRM, containing requirements ranging from the development of initiating events to the analysis of offsite radiological consequence. Based on feedback from pilot applications of the trial standard, a revised version of the standard was developed and balloted in 2020, with unanimous passage by the JCNRM and formal publication as an American National Standards Institute (ANSI) standard in 2021. The paper provides an overview of the standard and its development process. In addition, the paper reviews the latest status of U.S. Nuclear Regulatory Commission endorsement efforts through Regulatory Guide 1.247.

Fleming, K.↗

Advanced Features and Recent Developments in the WEC-Sim Open-Source Design Tool: Preprint

WEC-Sim is an open-source tool developed in Matlab/Simulink to simulate wave energy converters. WEC-Sim has been developed jointly by the National Renewable Energy Laboratory (NREL) and Sandia National Laboratories (SNL) since its initial release in 2014. WEC-Sim simulations are performed in the time domain based on the radiation and diffraction method using hydrodynamics coefficients derived from boundary element method (BEM)-based frequency-domain potential flow solvers (e.g., WAMIT, NEMOH, Capytaine, or ANSYS-AQWA). Within this level of modeling fidelity, WEC-Sim can handle floating body hydrodynamics, mechanical and electrical power generation methods, advanced control implementation, mooring systems, and other unique applications such as desalination. WEC-Sim functionalities are developed using Simulink blocks and Matlab functions. WEC-Sim is capabilities include simulation of different wave conditions, constraints and PTOs, body dynamics, mooring systems, among other advanced features such as nonlinear hydrodynamics, nonlinear PTOs, body-to-body interactions, etc.

hydrodynamics modeling↗

Flow in and Permeability of Porous Media for Turbulent Flow - 16117

The turbulent flow of an incompressible fluid in a porous medium with irregularly shaped grains is analyzed by adopting the k-ε model and making use of ANSYS. The effective permeability is dominantly in the longitudinal direction with negligibly small transverse component due to anisotropic arrangement of the grains. It is observed that the turbulent kinetic energy is large in the central region of pores, but the turbulent dissipation appears to be quite uniform. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Obtaining a Special Administrative License Exemption to Ship the West Valley Melter Box - 16327

In order to ship the West Valley Melter Package (WVMP) package to support its disposal, the U.S. Department of Energy (DOE) requested a one-time shipment license exemption from the U.S. Nuclear Regulatory Commission (NRC). Based on content, the previously constructed WVMP, containing residual borosilicate glass and Low Density Cellular Concrete was shown to require a Type B Package. Therefore, Title 10 of the United States Code of Federal Regulations, Part 71 (10 CFR 71) required that radioactive releases under Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC) would remain less than 1 E-7 A{sub 2}/hour and 1 E-1 A{sub 2}/week, respectively. Although the NRC normal method for demonstrating conformance to the 10CFR71 requirement is through performance of leak rate testing per American National Standards Institute (ANSI) standard N14.5, the WVMP does not contain a pressure vessel; therefore, performance of leak rate testing was not applicable/possible. Instead, an alternate method, adapted from DOE Handbook 3010 was used to show that the maximum allowed 10 CFR 71 radionuclide release rates for NCT and HAC would not be exceeded. Summarizing the effort, this paper provides the technical synopsis for requesting the exemption. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Content Addition and Testing for 2017 NRC Amendment for the Versa-Pac Type A Fissile Packaging - 18120

The Versa-Pac is a certified Type A, fissile packaging (USA NRC Certificate no. USA/9342/AF-96), often used as a waste package for one time use. Continuous content additions have improved the versatility of the package resulting in expanded uses for the nuclear industry. Principal design of the Versa-Pac packaging maintains the use of an inner liner positioned inside a 55-gallon (VP-55) or 110-gallon drum (VP-110), as used for decades in the transport and disposal of Type A fissile materials. The Versa-Pac is used directly or in conjunction with pails, pipe containers, poly bottles, and a variety of smaller containers, inserts and vessels. The Versa-Pac content developments assist the industry with site shutdown, decommissioning, and cleanup operations, allowing for removal, transport, and storage options for the residual waste materials. A 2017 NRC license amendment application adds new capabilities to the Versa-Pac that include a low 1.25 wt% {sup 235}U enriched uranium loading, small quantities of UF{sub 6} in the form of 1S and 2S cylinders, and a payload mass increase from 250 to 350 pounds for the Versa-Pac 55-gallon version (VP-55). The lower uranium enrichment loading of 1.25 wt% {sup 235}U allows for a higher uranium mass limit. The inclusion of 1S and 2S UF{sub 6} cylinders fills an industry need to remove and dispose of long-stored material during site cleanup activities. The content payload mass limit is increased for low contaminated or radioactive heavy metal contents, such as cut plant components and ingots. Regulatory certification of new contents is demonstrated by test, analysis, and a combination of these techniques to show that the design function of the Versa-Pac as a Type A fissile package remains unchanged. Specific UF{sub 6} criticality and thermal analysis upgrades are added to the design documentation to accommodate 1S and 2S UF{sub 6} cylinders. The parameters for criticality analyses are simplified, ensuring no credit for cylinder containment is required for HAC thus reducing the reliance on a single specific shoring device. An upgrade of the Versa-Pac thermal analysis demonstrates compliance with ANSI N14.1 UF{sub 6} cylinder thermal limits. A new drop test sequence for the VP-55 was performed, documenting the addition of the 350-pound payload and demonstrating the VP-55 maintains structural integrity during the NCT and HAC drop, crush and puncture events as defined by 10 CFR 71. The testing and analyses show no loss of the material occurs and the package maintains the material in a subcritical state. This paper traces the engineering and licensing efforts involved in expanding the Versa-Pac's versatility for the 2017 NRC amendment application; summarizing the new testing and results for the payload increase, the content addition criticality and thermal evaluations, and content description for the Versa-Pac Safety Analysis Report. The NRC license amendment is scheduled for Winter 2017. Testing and supporting analyses reported in the Versa-Pac Safety Analysis Report demonstrates the content developments met the performance requirements of 10 CFR 71 for a Type A fissile package. A revision of the NRC Certificate of Compliance is expected in Spring 2018. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Moving Beyond Traditional Tornado Missile Impact Analysis

On March 11, 2011 a 9.0 magnitude earthquake occurred off the coast of Japan creating a tsunami, which together severely damaged northeastern Japan including the Fukushima Dai-ichi nuclear power plant site. In response to this event, the United States (U.S.) Nuclear Regulatory Commission (NRC) issued Order EA-12-049, 'Issuance of Order to Modify Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events'. In order to meet the guidance of the NRC Order EA-12- 049, the a central U.S. nuclear generating station (Plant) was required to demonstrate that the currently installed Borated Water Storage Tank (BWST), Refueling Water Tank (RWT), and Qualified Condensate Storage Tank (QCST) could provide the necessary capacity following a Beyond Design Basis External Event (BDBEE). These tanks were not originally designed to meet current site design basis requirements for tornado wind and missile impacts. To save the Plant from having to build costly new storage tanks off site, the existing tanks were credited as robust for tornado missile impact through advanced finite element impact analyses using the ANSYS Autodyn solver and fluid-structure interaction (FSI). This procedure is the first known NRC acceptance of advanced analysis and FSI for to satisfy the FLEX tornado missile impact criteria. Extensive verification and validation (V and V) was performed to verify perforation of steel plates which included FSI. These V and V analyses were confirmed against large databases of experimental and theoretical results to ensure accurate modeling and results. The use of this advanced analytical method benefitted the Plant in several ways. A new storage tank was not required to be built, saving not only initial construction costs but also ongoing maintenance costs and monitoring activities. By determining specific vulnerable components of each tank a set of modifications and protection schemes were developed. These modifications were much quicker to implement saving the Plant on downtime costs. (authors)

42 ENGINEERING↗

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)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Using CFD tools to model separate effects of HTR air ingress accident phenomena - 18508

The air ingress phenomena under the HTR conditions represent the important area of nuclear safety. Best practice guidelines for the use of CFD in nuclear safety applications are mainly based on water cooled reactors while in HTR the gas is used as a coolant. The ambition of this paper is to define the best recommendations and description of needed settings of CFD models solving the separate effects of air ingress accident phenomena. The JAERI experiments deal with phenomena of air ingress during the early stages of the accident. Experiments investigate the basic features of the flow behavior of the gas mixture and regard the phenomena of the molecular diffusion and the natural convection of the gas mixture with the graphite oxidation reaction. The separate effects simulations of air ingress accident were performed using CFD code ANSYS Fluent. Whole graphite oxidation process including all three main chemical reactions was successfully modeled. The new findings and useful information were gathered to be extrapolated to the integral effects level for the following phase. The role of the computational mesh and all important input parameters were discussed in detail and many in sensitivity studies were performed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Multi-Physics Coupling to Model a RbCl-RbCl-Ga Target Stack

A computational model has been developed to simulate the performance of the Los Alamos Isotope Production Facility's (IPF) routine production target configuration, which consists of two Inconel-encapsulated RbCl salt targets and one niobium-encapsulated gallium target separated by water cooling channels. The methodology involves iteratively coupling particle transport and thermal hydraulic codes to determine the time dependent behavior of the target. Phase and density changes seen in targets are known to impact beam penetration and distribution. In addition, evidence of energy distributions seen in experimental radionuclide yield measurements at IPF do not match those seen in previous MCNP predictions. Density reduction as a result of heat deposition affects proton energy loss along the particle's track, impacting both radionuclide yields and target design. Accurate modeling of the thermal behavior of these multi-phase targets and the corresponding impact on beam behavior will significantly improve our understanding of incident and exit proton energy distributions and resulting radionuclide yields in downstream targets. Thermal hydraulic modeling, including computational fluid dynamics (CFD) and heat transfer, was performed with ANSYS CFX while particle transport was performed with the Monte Carlo N-Particle code (MCNP). (authors)

07 ISOTOPE AND RADIATION SOURCES↗