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At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

On the influence of the power plant operational history on the inventory and the uncertainties of radionuclides relevant for the final disposal of PWR spent fuel - 15149

A mandatory requisite for official granting safe final disposal approval of SNF is the comprehensive specification and declaration of the nuclear inventory in SNF by the NPP-operators. The whole radiochemical characterization of each spent fuel assembly hides many difficulties such as very high radiation burden, destruction method, the inhomogeneous burn-up of fuel assembly, very high cost. Thus, non-destructive characterization is usually applied on industrial scale. In this case only several nuclides can be measured, and the rest is calculated using simulations and typical irradiation history. Like any other German rad-waste acceptance requirements all these declarations will have to be checked and verified by an independent inspector. In this verification process not only values of radionuclide (RN) activities are important, but also their uncertainties. The simple simulation using burn-up calculations based on typical and generic reactor operational parameters do not answer the question about possible uncertainties observed in real reactor operations. At the same time, the details of irradiation history are often unknown, which complicates the assessment of declared RN inventories. Therefore, the development of dedicated generic proof tools, which can be applied to different types of waste, often of unknown origin and history is worth a serious consideration. Here, we have compiled a set of burn-up calculations considering the operational history of 339 published or anonymized real PWR fuel assemblies. This gives us the possibility to calculate the realistic variation or spectrum of RN inventories. These histories were used as a basis for 'SRP analysis', to provide information about the range of the values of the associated secondary reactor parameters (SRP's), i.e. initial enrichment (IE), specific power (SP), irradiation time (IT), downtime (DT), boric acid concentration (BA), moderator and fuel density (MD and FD) and temperatures (MT and MT) used as input parameters for the simulation models. Then, burn-up calculation codes provide the anticipated results for the entire nuclear inventory such as characteristic activity values and the corresponding minimum and maximum limits as a function of the fuel burn-up. These results are being validated with the experimental data from various publications and the online database - SFCOMPO and subsequently transferred and summarized into our own database. Ultimately, this database will provide the basis of proof tools for the operator independent verification of the declared nuclear inventories for the quality and safety assessment of a radioactive waste compound and bounding values for certain RN as a function of burn-up. The RN inventory can hence be verified without taking credit of the whole of the operational history of the SNF. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Effect of sodium / boron mole ratio on the cement solidification for sodium borate waste produced in PWR plants - 15261

A cement solidification process for treating sodium borate waste produced in pressurized water reactor (PWR) plants was studied. The process was developed to obtain a high mechanical strength of more than 10 MPa which can satisfy various countries disposal regulations. To obtain high volume reduction and high mechanical strength of the solidified product, simulated concentrated borate liquid waste was dehydrated and powdered by using a wiped film evaporator. To investigate the effect of the sodium / boron (Na/B) mole ratio on the strength of cement solidified waste, simulated concentrated borate liquid waste with Na/B mole ratio between 0.21 and 0.85 were used. Solidified cement containing powdered waste with a Na/B mole ratio 0.85 showed to excellent operability and uniaxial compressive strength. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Development of Cement Solidification Technique for Sodium Borate Waste Produced in PWR Plants - 16061

A cement solidification process for treating sodium borate waste produced in pressurized water reactor (PWR) plants was studied. It is known that the setting and hardening reaction of cement is retarded by the presence of boron compounds. This technique was developed to obtain high uniaxial compressive strength more than 10 MPa which can satisfy various countries disposal regulations. As examples of disposal regulations, France requires 8 MPa for homogeneous solidified products. The sodium/ boron (Na/B) ratio was controlled to about 1 by adding NaOH to simulated liquid waste for the purpose of improving the workablility and mechanical strength. To obtain high volume reduction, simulated liquid waste was dehydrated and powdered by using a wiped film evaporator. Ordinary portland cement (OPC) was used for the solidification. The viscosity of cement mixture was controlled by the water/cement ratio. The resulting viscosity became less than 25 dPa.s, and it was possible to continuously mix the mixture. Solidified cement with containing 73 kg/m{sup 3} of boron had a uniaxial compressive strength of more than 10 MPa. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Comparisons of Peak Cladding Temperature and Effective Thermal Conductivity: CFD Simulation on PWR Spent Fuel Assemblies - 16278

When spent fuel assemblies from the reactor of nuclear power plants (NPPs) are transported or stored, the assemblies are exposed to a variety of environments that can affect the peak cladding temperature of spent fuel assemblies. There are four methods to calculate the peak cladding temperature of spent fuel assemblies in a cask: Manteufel and Todreas's two-region model, Bahney Lotz's effective thermal conductivity model, Wooton-Epstein correlation, and computational fluid dynamics (CFD) simulation. The CFD simulation based on the FLUENT code is used to calculate the peak cladding temperature and effective thermal conductivity of a spent fuel assemblies used in Republic of Korea NPPs: 14x14, 16x16 ACE7, 16x16 PLUS7 and 17x17 PWR spent fuel assemblies. The CFD simulation results show that the effective thermal conductivity of the 16x16 ACE7 was more conservative than those for the other assemblies. Therefore the effective thermal conductivity using 16x16 ACE7 is used for thermal evaluation of the KORAD's developing cask which can load 14x14, 16x16 ACE7, 16x16 PLUS7 and 17x17 spent fuel assembly. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Developmental study of advanced HIS design method for digital I and C+HMIT of PWR plant - 277

A new conceptual frame of how to design and validate a digital HIS (human interface system) on an innovative numerical simulation basis is proposed for the support of plant operators' supervisory control of various types of automated complex NPPs (nuclear power plants). The proposed conceptual framework utilizes the object-oriented software for plant DiD (defense-in depth) risk monitor with the combination of nuclear reactor accident simulation by an advanced nuclear safety analysis code RELAP5/MOD4. The conceptual frame proposed in this paper will be applied for an example practice for the SBLOCA (small break loss of coolant accident) case of passive safety PWR (pressurized water reactor) AP1000. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

GSI 191 Closure Issue for EDF PWR Fleet: The Fuel Assembly Test Results

EDF has re started recirculation studies within the international GSI 191 framework by 2002, when EDF decided to join the NEI sump strainers task force. All the EDF PWR fleet have been upgraded, with major strainer modifications, according to the NEI 04-07 guideline. The so-called chemical effects have been carefully studied by EDF and demonstration have been released that there are no chemical effects on EDF fleet, owing to the nature of our debris. The study of strainer downstream effects started by 2009, with decoupling Downstream Sump Debris Source Term (DST) and by the end of 2016, all the equipment have been successfully qualified. At a previous ANS meeting,it was said that core fuel tests were scheduled. They have started at the beginning of 2014 and they have been closed by mid 2016 for AREVA Fuel type at ERLANGEN facility (Germany). The present presentation is aimed at describing the test facility, the test specifications and the main results. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Sensitivity study of PWR upper vessel flow and phenomena using RELAP5-3D

During normal pressurized water reactor (PWR) operation, water enters the reactor pressure vessel (RPV) from the cold leg and exits through the hot leg. While most of this water flows through the downcomer and reactor core, a number of other flow paths exist, including the core baffle/bypass and the upper plenum spray. The upper plenum sprays normally account for a small fraction (less than 10%) of the total inlet flow, and provide cooling to the upper vessel dome and instrumentation/control rod channels. In the event of a loss-of-coolant accident (LOCA), especially one with core blockage, this alternative flow path may play a key role in maintaining core coolability. Because of the possible safety implications of the upper vessel flow, it is important to ensure the models used to simulate this region of the vessel are accurate. In this study, the model was altered to better reflect the internal flow path of the upper head and this was compared to the reference model. The simulation of interest is a hot leg double-ended guillotine (HL DEG) break followed later by a full blockage of the core and core bypass. In this scenario, the coolant can only reach the core through the upper spray bypass, so its impact on core coolability can be clearly seen. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Reactivity impact of accident tolerant claddings in an equilibrium PWR core

The US DoE's Accident Tolerant Fuels (ATF) program aims to establish a process to evaluate the feasibility of several ATF concepts. The two main aspects of such a process are to evaluate the 3-D 'whole core' neutronics and fuel (thermal-mechanical) performance. Determining the best ATF candidate usually involves a compromise between these two performance aspects. The work presented in this paper focuses solely on the 3-D neutronics, looking specifically at the end of cycle (EOC) reactivity, and upon the establishment of a realistic and modern equilibrium PWR core to use as a test-bed. Both the lattice physics code CASMO-4 and the 3-D nodal simulator NESTLE were used in this study. Eventually, operational power histories are provided to our collaborating team to carry out fuel performance evaluations to be presented in a future article. Neutronically, a goal of a prospective ATF concept is to at least match the fuel cycle length of a core with standard Zircaloy/UO{sub 2} fuel, without having to significantly change the design of the fuel bundle; meaning, without changing pin pitch or outer clad diameter. Such a 'seamless' transition to another cladding type would greatly reduce the burden of all the nuclear energy stakeholders (regulators, vendors, utilities), from both an analysis and financial perspective. For cladding materials that are less transparent to neutrons, which for this study included all the ATF concepts, the two degrees of freedom with respect to fuel fabrication are the U-235 enrichment and cladding thickness. The latter involved changing only the fuel radius and inner clad radius while keeping the gap and outer clad radius constant. The thinnest cladding value was set to a generous estimate of the minimum iron-chromium-aluminum (FeCrAl) thickness that can be manufactured. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

An Unsteady State Gas Transport Model in Matrix Laboratory (MATLAB) for PWR Gas Management

Reactor Coolant System (RCS) hydrogen gas management can be a challenge at Pressurized Water Reactors (PWRs) during shutdown, startup, and steady state operation. Poor hydrogen control can result in higher general corrosion rates for RCS components and increases in critical path time for shutdown or startup. Also, performance indicator penalties may be accrued by entry into Electric Power Research Institute (EPRI) Action Level conditions or by accrual of Chemistry Effectiveness Index (CEI) penalties established by the Institute of Nuclear Power Operations (INPO). Furthermore, poor hydrogen control may increase industrial safety risk due to the creation of an explosive environment. Part of the hydrogen management difficulty arises from the interaction of several phenomena that require complex simulations to fully understand. While hydrogen gas transport models have been developed, the application of these models is limited by the computational intensity required to accurately model the dynamism inherent in plant shutdown or startup. Subsequently, a gas transport model is developed in a suitable computational program to meet these needs and provide input to an optimized RCS hydrogen management strategy for all operational modes. The gas transport model is used to model RCS hydrogen behavior for an example PWR during an End of Cycle (EOC) Forced Outage (FO) startup, a refueling outage (RFO) shutdown, and during nominal steady state operation. The computational software selected is Matrix Laboratory (MATLAB{sup R})

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Heterogeneous Neutron Leakage Model for PWR Pin-by-pin Calculation

In the last few decades, the nodal diffusion calculation accompanied with the assembly homogenization has been widely employed. The single assembly calculation with the reflective boundary condition (BC) would be carried out to determine the assembly-homogenized parameters. Because of the reflective BC, no axial and/or radial leakage rates have been taken into accounted for the lattice calculation. To represent the global influence of the reactor core on an assembly calculation, one mostly frequently uses a procedure called leakage model to approximately take into account the absorption due to the macroscopic structure of the physical flux in a heterogeneous situation. For the assembly homogenization, this absorption is introduced as the product of the leakage coefficient and the buckling obtained for an equivalent homogeneous medium using the well-known homogeneous B1 theory. Clearly, this is an imperfect representation of leakages in a heterogeneous medium. However, the errors introduced by the foregoing procedure are small for the whole? -assembly homogenization in a PWR. When it comes to the pin-by-pin calculation, it is not the case. The leakage rate? will be affected by the neutron streaming effect caused by the heterogeneity of the assembly. Especially when poison pins or empty channels are present, the streaming effect becomes important. The errors introduced appear particularly visible. The heterogeneous leakage model? which assumes that the neutron flux is variational and the leakage rate is not uniform over the lattice pitch is a more appropriate choice. When the heterogeneous leakage model is applied in the lattice calculation for the pin-cell homogenization, the pin-homogenized problem would be changed compared with the original assembly problem without any homogenization technique. So, the Cooperation of the heterogeneous leakage model with the pin-cell homogenization theory is unsettled. In this paper, both homogenous and heterogeneous leakage models are evaluated in pin-by-pin calculation. Their implements used in pin-cell homogenization of the lattice calculation are studied. The diffusion coefficient obtained from transport cross-section and the space-dependent leakage coefficient together with the two leakage models are analyzed for the practical reactor-core applications.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Verification of AZTRAN v1.2 code with a PWR-MOX benchmark problem

The AZTLAN platform project is an effort focused on developing a modeling Mexican platform to design and analyze nuclear reactors. It is a national initiative lead by the National Institute of Nuclear Research (ININ), in collaboration with the main Mexico's public universities: the National Polytechnic Institute (IPN), the National Autonomous University of Mexico (UNAM) and the Autonomous Metropolitan University (UAM). The AZTRAN code is a module of AZTLAN that solves the 3D neutron transport equation using the discrete ordinates (SN) method and the nodal scheme RTN-0 on steady state and Cartesian geometry. AZTRAN works with cross sections generated externally by other code. It provides a solution to the eigenvalue problem and determine the neutron distribution for a specific medium at desired conditions. On the other hand, it is important to mention that the Working Group 4 (users group) of AZTLAN platform is responsible of verify and validate the codes development, in order to give feedback to the group of developers: neutronics and thermal-hydraulics. In this work, the AZTRAN v1.2 performance was studied for a PWR-MOX lattice, using as example the Yamamoto's benchmark for LWR fuels, comparing AZTRAN v1.2 results again those obtained with the well-known neutronic code SERPENT-2 (v1.27), and some results from an article presented at PHYSOR-2002. Results from AZTRAN for a BWR fuel lattice is presented in a companion paper in this meeting.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Modeling the Transient Reactor Test Loop in TRACE for PWR Critical Heat Flux Benchmarking

Understanding transient critical heat flux (CHF) resulting from a Reactivity Insertion Accident (RIA) in nuclear systems is an extremely undeveloped field. Though many experiments have been done for CHF, remarkably few have been done for rods at PWR operating temperature and pressure (300 deg. C, 2200 psia) with a short time constant ramp in power. This leads to many concerns regarding understanding how RIA scenarios develop and evolve in commercial nuclear reactors. With a limited understanding of the phenomenology governing transient CHF, there is little hope in being able to accurately model these scenarios using today's thermal hydraulic codes. The Accident Tolerant Fuels (ATF) Program established by the Department of Energy (DOE) is seeking out new fuel and cladding systems that can be used to reduce the probability and consequences of severe accidents. To better understand this phenomenon, the DOE is restarting the Transient Reactor Test (TREAT) Facility at the Idaho National Laboratory (INL). TREAT will be able to provide representative neutronic boundary conditions for testing coupled thermal hydraulics on the Transient Water Environment Recirculating Loop (TWERL) to study transient CHF with the best representation possible. To support this task, Oregon State University (OSU) is designing and building a representative flow loop that will be able to mimic the conditions of TWERL (sans the neutronic boundary conditions) to create a benchmark study for existing lumped parameter thermal-hydraulic codes such as RELAP5 and TRACE. This flow loop, the Transient Reactor Test Facility (TRTL), will support the ATF program by providing much needed NQA-1 data for transient CHF at operating temperature and pressure.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Computational Study of Droplet Dynamics in Dispersed Flow Film Boiling Regime in PWR Geometries

Dispersed flow film boiling (DFFB) regime is exhibited during low-flooding rate conditions following the postulated Loss-of-Coolant Accident (LOCA) in nuclear reactor flow channels. LOCA is a type of design-basis accident scenario, which is crucial for reactor design and licensing. Accurate models to predict the heat transfer in DFFB regime is critical for the reactor to maintain safety limits. The nominal operating linear heat rate is also set based on the maximum linear heat rate determined for a LOCA transient. It is understood that the behavior of entrained droplets in DFFB regime greatly affects the heat transfer by evaporating and desuperheating the surrounding steam. Furthermore, interaction with flow structures also aids in increasing the evaporation rate of entrained droplets. In order to quantify this enhancement in heat transfer, the size of droplets downstream of flow structures (Spacer Grids and Mixing Vanes) is of crucial importance. Thermal hydraulic analysis codes like COBRA-TF, CATHARE-3 and adopt a three-field equation model that allows them to model the droplet field separately. However, to ensure the accuracy of these field equations, various correlations and physical models are required. This paper proposes a method for characterizing the mean Sauter Diameter of droplets across spacer-grids and mixing vanes in PWR sub-channels using PHASTA - a high fidelity flow solver with multiphase capabilities using the level-set method. The PHASTA code is a finite element based computational fluid dynamics code designed for solving compressible and incompressible turbulent flows. It has been demonstrated to be massively scalable (up to 786,432 cores), which is crucial for simulating large realistic domains. The code incorporates the level-set method for interface tracking of the gas-liquid interface.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Activation Neutronics for a Swiss PWR: Simulations and Validation - Paper 102

In Switzerland 40% of the electricity generation is produced by nuclear power. With all four NPPs being already beyond their 30. year of operation, Nagra in collaboration with the utilities periodically contributes to the Swiss NPP decommissioning cost estimation studies. The neutron induced activation of structures and components of the NPPs is of particular relevance for the planning of the dismantling actions, the waste-packaging concept, the waste-management considerations and, therefore, the estimation of decommissioning costs. In this context, Nagra has developed a calculation sequence in order to perform state-of-the-art neutron transport and activation calculations. In this paper, aspects relevant to the neutron transport calculations for estimating the residual radioactivity within a Swiss PWR are presented. More specifically, the MCNP5 modeling approach together with the use of the ADVANTG-hybrid code for accelerating the variance reduction efforts is outlined. Additionally, the validation of the neutron transport calculations with in-situ full-cycle foil activation campaigns is discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Extensive sensitivity study of PWR fuel assay data - 14301

Accurate calculations of the isotopic composition of spent nuclear fuel play a fundamental role in several parts of nuclear energy. An important example is the use of burnup credit in criticality safety analysis, where the results gained from depletion calculations are used to determine the fuel composition in subsequent criticality calculations. The work presented here is a sensitivity study for depletion calculations of high-burnup fuel samples from the Vandellos II PWR that entails an unusually detailed evaluation of assay data. In addition to investigating the impacts of experimental uncertainties on the calculated isotopic composition, the consequences of various modelling approaches are studied to give quantitative estimations of the impact on accuracy. Furthermore, the reactivity of the calculated fuel composition is tracked, because this is the integral characteristic of the irradiated fuel that one is actually interested in for burnup credit applications. While the sensitivity study is performed only for one of the nine experimental samples of the Vandellos II assay data, the lessons learned lead to evaluation models with different levels of model detail for all fuel samples. The results show that the discrepancies between the measured and calculated isotopic compositions are considerably larger than the uncertainties inherent in the measurement and the calculation method would indicate. This is not unexpected, since large discrepancies are a common result of assay data recalculations. However, the unusually extensive sensitivity study performed here allows to exclude some possible sources of error: since the influence of experimental uncertainties like fuel assembly geometry and reactor parameters on the isotopic composition is shown to be small, one valuable conclusion of this study is to focus more on the measurement uncertainties of the chemical assay itself, and the nuclear data that the calculation is based upon. The study also demonstrates that using the reactivity of the depleted fuel as an additional comparison point between measurement and calculation can provide some orientation regarding the importance of the manifold influences on calculation results. Tracking this integral characteristic of the depleted fuel is very helpful in focusing on the actual impact of the depletion calculation uncertainties on criticality safety. (author)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Neutronic analysis of a PWR-type SMR core using duplex ThO{sub 2}-UO{sub 2} in TRISO fuel particles

Currently, studies on small modular reactors (SMRs) present an important development due to the potential they represent in terms of safety, operational flexibility, economy, and non-energy applications. Furthermore, there is increasing interest in studying the use of thorium as fuel, as an effective way to solve problems such as the shortage of uranium reserves, reduction of nuclear waste and nuclear proliferation. Also, the use of thorium combined with highly enriched uranium TRISO particles has been studied, showing proper performance. In this work, the concept of ThO{sub 2}-UO{sub 2} duplex fuel is used, for the core configuration of a PWR type SMR that uses TRISO fuel, designed to achieve extended fuel cycles. Three distribution cases of ThO{sub 2} and UO{sub 2} in TRISO particles inside the fuel rods are compared. First, particles composed by a homogeneous mixture of ThO{sub 2} and UO{sub 2} are distributed inside of fuel elements. Second, the fuel zone of the fuel elements is divided into two radially, an internal one where the TRISO particles composed of ThO{sub 2} are distributed and the external one where the TRISO particles composed of UO{sub 2} are distributed. The third case is like the previous one, except that the particles that contain UO{sub 2} are distributed in the inner zone and those that contain ThO{sub 2} in the outer zone of the fuel elements. The comparison of the cases is carried out in terms of cycle main isotopes' mass transmutation, power distribution and temperature reactivity coefficients. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Microstructural Effects on SCC Initiation PWR Primary Water Cold-Worked Alloy 600

SCC initiation behavior of one mill annealed alloy 600 plate heat was investigated in simulated PWR primary water under constant load at yield stress with in-situ direct current potential drop (DCPD) monitoring for crack initiation. Twelve specimens were tested at similar cold work levels among which three showed much shorter SCC initiation times (<400 hrs) than the others (>1200 hrs). Post-test examinations revealed that these three specimens all feature an inhomogeneous microstructure where the primary crack always nucleated along the boundary of large elongated grains protruding normally into the gauge. In contrast, such microstructure was either not observed or did not extend deep enough into the gauge in the other specimens exhibiting ~3-6X longer initiation times. In order to better understand the role of this microstructural inhomogeneity in SCC initiation, high-resolution microscopy was performed to compare carbide morphology and strain distribution between the long grains and normal grains, and their potential effects on SCC initiation are discussed in this paper.

Alloy 600↗
DOE OSTI2017

Grain boundary damage evolution and SCC initiation of cold-worked alloy 690 in simulated PWR primary water

Long-term grain boundary (GB) damage evolution and stress corrosion crack initiation in alloy 690 are being investigated by constant load tensile testing in high-temperature, simulated PWR primary water. Six commercial alloy 690 heats are being tested in various cold work conditions loaded at their yield stress. This paper reviews the basic test approach and detailed characterizations performed on selected specimens after an exposure time of ~1 year. Intergranular crack nucleation was observed under constant stress in certain highly cold-worked (CW) alloy 690 heats and was found to be associated with the formation of GB cavities. Somewhat surprisingly, the heats most susceptible to cavity formation and crack nucleation were thermally treated materials with most uniform coverage of small GB carbides. Microstructure, % cold work and applied stress comparisons are made among the alloy 690 heats to better understand the factors influencing GB cavity formation and crack initiation.

Alloy 690↗