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At least 199 records · Page 11
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Simulation and control of a passively cooled small modular reactor Samet - 142

This work regulates a small modular reactor (SMR) with traditional control methodologies and analyzes its response to typical disturbances. An analytical model for the passively cooled NuScale SMR is developed to represent the essential dynamics. The MATLAB/Simulink based model consists of reactor core, steam generator, pressurizer, hot leg riser and downcomer, and turbine representations. The point kinetics equations with a single combined neutron precursor group and the lumped parameter representations with single-phase natural circulation account for the neutronics and thermohydraulics in the reactor core. A lumped parameter, moving-boundary approach forms the basis of the nonlinear model adopted for the once-through helical-coil heat exchangers in which boundaries between regions of different fluid states (i.e., subcooled, boiling, and superheated) vary over time. A proportional-integral-derivative controller employing a sliding-average temperature strategy commonly used for pressurized water reactors is implemented. The responses of the SMR to two different disturbances are analyzed and discussed in detail. For the first scenario, a 5% step increase in the load is applied to the system, which results in a change in the steam valve opening. For comparison, two different simulations under this same disturbance are run: one including and one excluding the control systems. The second scenario tests the effectiveness of the control system to increase (or decrease) the reactor thermal power to a certain level within a desired time period. The simulation results show that important system variables are kept at the desired values by the proposed control strategy. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Assessment of boric acid accumulation in a reactor core blocked by debris

To examine the potential for boron precipitation in the core during sump recirculation when latent debris has formed a blockage at the core inlet, TRACE simulations of long term cooling following a loss-of-coolant accident (LOCA) in a Westinghouse four loop pressurized water reactor (PWR) were performed. The TRAC/RELAP Advanced Computational Engine (TRACE) was modified to account for the higher fluid densities and viscosities that occur when the coolant contains boric acid. Previous versions of TRACE calculated the transport of boron in the RCS, but assumed coolant properties to be the same as pure water. This was appropriate for very low concentrations of boric acid, but not when the concentrations approach those where precipitation is possible. These modifications allow TRACE to simulate natural circulation caused by a concentration difference in addition to the buoyancy induced by temperature differences. The results of the TRACE calculations suggest that there is margin to boric acid precipitation during long term cooling following a large break LOCA even if debris significantly affects flow at the core inlet. Significant blockages at either the lower core plate or bottom fuel nozzles did not result in boric acid concentrations likely to cause precipitation in the core. The calculations do show that precipitation can occur in the upper head where coolant can become trapped and in the steam generators where entrained fluid can evaporate and deposit the boron carried by the droplets. Of particular importance, is that the boric acid concentration and coolant temperature in the lower plenum and regions near debris blockage remained well below precipitation levels supporting a conclusion that precipitation is not likely to make the blockage more severe. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Dismantling of the Bohunice V1 Reactor Coolant System - 20011

About 2.5 kilometers from the village of Jaslovske Bohunice in the Trnava District in western Slovakia at the Bohunice V1 nuclear power plant, Westinghouse and VUJE are working on the dismantling of two VVER-440 units. The consortium was awarded this contract in October 2017 from Jadrova a vyradovacia spolocnost (JAVYS) and has begun onsite work to dismantle the reactor coolant systems of both units that operated at the plant. The Bohunice V1 plant is composed of two V-230 type VVER 440 units (440 MWe each). The first unit ceased operation on December 31, 2006, and the second on December 31, 2008, after 28 years of operation. The project strategy relies on the early release of some plant areas to allow the erection of new facilities to enable a fast and safe dismantling. These up-front activities play a crucial role within the whole project. Since schedule and ALARA are two key elements to be considered, it has been decided to separate the dry cutting activities from the ones occurring underwater. Two Wet Cutting Workshops will allow the segmentation and packaging of the high activity components. To take advantage of the large space available in the turbine building, a Dry Cutting Workshop has been built for segmenting the twelve steam generators of both units. Two more Dry Cutting Workshops (one per unit) have been installed in the reactor building for the fragmentation of the remaining large components and auxiliary systems components. This paper describes all current preparatory works as well as the on-going segmentation and packaging activities. Interesting lessons learned are also provided. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Predicting Rupture Pressure and Leak Rate of Alloy 600 Tubes with ODSCC and Comparing with Test Results (Technical Letter Report)

This report summarizes structural integrity work performed for the U.S. NRC that predicts the rupture pressure and leak rate of flawed steam generator (SG) tubes. The work includes pressure testing of straight tubes with lab-induced outer diameter stress corrosion cracks (ODSCC). Existing models for predicting rupture pressure and leak rate of straight tubes with well-defined, rectangular, electro-discharge-machined (EDM) flaws were modified and used to predict the measured rupture pressures and leak rates from straight tubes with ODSCC. Models previously developed for EDM notches in straight tubes were modified for ODSCC by using the equivalent rectangular crack (ERC) method.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Sensitivity Analysis of OECD ATLAS A5.1 Test using MARS-KS code

The ATLAS was designed to model a reduced-height primary system of a typical APR1400 which was developed by the Korean Nuclear Industry. The ATLAS has the 1/2-height, 1/144-area and 1/288-volume scales for the APR 1400 and it is scaled for full pressure and temperature conditions. A benchmark test, test A5.1, was performed to analyze and quantify the extent of scaling distortion in the ATLAS, as a part of a OECD-ATLAS project. The SB-CL-32 test which was conducted using the Large Scale Test Facility (LSTF) of the Rig-of-Safety Assessment-V (ROSA-V) Program was selected as the counterpart test to address the scaling issue. The detailed description of the LSTF test facility and the SB-CL-32 test can be found in the literature. The objective of the A5.1 test is to investigate the thermal hydraulic phenomena during cold leg small break LOCA with steam generator secondary side depressurization as the Accident Management action. In addition to that, the scaling effect analysis of ATLAS was performed in the several viewpoints. In this study, sensitivity analysis of A5.1 test was performed using MARS-KS code. Several experimental conditions of A5.1 test were varied to study effects on the reactor coolant system during the SBLOCA transient. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Renewable Thermal Energy Systems Designed for Industrial Process Solutions in Multiple Industries

The need for renewable heat in industry is vital for the next decade and beyond. Industrial decarbonization is a key area that must be accelerated, to foster the removal of fossil fuels from the provision of heat, especially at low temperatures. This paper looks at the development and results of case studies for understanding the economics and potential for renewable thermal energy systems (RTES), particularly in hybrid configurations to provide industrial process heat (IPH). For the case studies, these include non-concentrating e.g., heat pumps, and concentrating collectors e.g., parabolic trough collectors and direct steam generation (DSG)-linear Fresnel collectors (LFCs). The results show that a levelized cost of heat (LCOH) of $6-$8 per million British Thermal Units (MMBTU) is possible, depending on the direct normal irradiance (DNI) and the system sizing e.g., to increase the solar fractions. In Arizona for example, with a DNI of 7.36 kWh/m2/day, the base case for the DSG-LFC system with 6hrs of thermal energy storage could potentially meet a 1 MWth load 80% of the year.

concentrated solar thermal↗
DOE OSTI2019

Lessons Learned from Most Recent Chemical Decontamination for Decommissioning Projects - 19280

In 2017, Westinghouse performed a full system chemical decontamination of Units 1 and 2 of the Bohunice V1 Nuclear Power Plant. Bohunice V1 is a VVER 440 design with six loops. The application was performed two loops at a time, with the reactor core bypassed. The Pressurizer and Pressurizer Relief Tank (Bubble Tank) were decontaminated as well, in a separate application. This was a recovery decontamination that had been attempted by another company prior to Westinghouse. Several modifications to equipment and design of process were required to make the decontamination successful. The project was initiated to perform pre-dismantling chemical decontamination of primary circuits of both V1 units using Decontamination for Decommissioning (DfD) to meet the following objectives (see References 1 and 2 for process descriptions of DfD): - Remove activity from the primary circuit components; - Reduce dose rates around piping and components; - Simplify access to the installations, making it possible to use hands-on techniques for dismantling, rather than the more expensive use of robotics or manipulators; - Minimize the potential for spreading contamination during decommissioning activities - Reduce the activity of components that they may be disposed of at a lower classification, therefore more economical, waste disposal category. Several techniques were employed that were unique, or first of a kind, to aid in the decontamination, including resin regeneration and bypass of the reactor core, while still providing enough flow to to allow sufficient dispersal through all the steam generator tubes. Equipment that was already available on site was evaluated and, when possible, combined with Westinghouse equipment to minimize application cost and schedule. The lessons learned from the first unit (Unit 2) were applied to the second unit (Unit 1) application. The partial relocation of equipment allowed for the most efficient schedule to complete the second decontamination. The paper will describe in more detail the results and lessons learned from the full system chemical decontaminations performed, as well as updates on another recent decontamination project. (authors)

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

A New Graphical Dismantling Process Simulation Technology for Flexible Planning of Nuclear Facility

The dismantling of highly radioactive components of the nuclear facility is one of the most difficult activities in decommissioning. In the planning stage, the dismantling scenario including dismantling equipment and processes has a significant effect on the cost and safety of the whole decommissioning. High level radiation of the dismantling object such as a reactor pressure vessel (RPV) and its internals restricts access by human workers, and makes an accident during the dismantling process more difficult to deal with. Decommissioning of nuclear facilities demand a large scale of budget and high level of safety, and therefore the preliminary verification of dismantling scenarios by graphical simulation has special importance at planning stage. In recent years, the manufacturing industry has aggressively adopted the virtual manufacturing technology to verify the facilities and processes with high efficiency. The virtual manufacturing technology was beginning to be adopted in the area of the nuclear facility decommissioning. The dismantling process of the Rotary Specimen Rack (RSR) in the Korea Research Reactor-1 and 2 was established and it was also applied to replace the steam generators on time and on budget in San Onofre Nuclear Generating Station. Current virtual manufacturing technologies are not sufficient for the area of nuclear facility decommissioning because the object segmenting function absolutely required for graphical simulations over various dismantling scenarios is unavailable. This paper proposes a new technology for graphical decommissioning simulator which is able to simulate various dismantling scenarios over nuclear facilities including repetitive object cuttings of heavy and complex structures by applying flexible cutting algorithm.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Renewable Thermal Energy Systems Designed for Industrial Process Solutions in Multiple Industries

Industrial decarbonization is a key area that must be accelerated, to foster the removal of fossil fuels from the provision of heat, especially at low temperatures less than 300 degrees C. This paper looks at the results of two case studies for understanding the economics and potential for renewable thermal energy systems (RTES), particularly in hybrid configurations to provide industrial process heat (IPH). The first case study looks at heat pumps for district heating, and the second, the use of linear Fresnel collectors (LFCs) coupled with phase change material (PCM) thermal energy storage (TES) for direct steam generation (DSG). Using district heat as an input for the heat pump, three cases were run harvesting energy from ambient water (5 degrees C), sewage water (20 degrees C), and a solar collector (35 degrees C). Accounting for elevated costs of infrastructure for each heat source, the levelized cost of heat (LCOH) of the first case study ranged from $4-$15 per million British Thermal Units (MMBTU). For the second case study modeling LFCs with PCM and TES, the results show that a LCOH of $9-$15 per MMBTU is possible, depending on the direct normal irradiance.

concentrated solar thermal↗
DOE OSTI2022

Renewable Thermal Energy Systems Designed for Industrial Process Solutions in Multiple Industries: Preprint

Industrial decarbonization is a key area that must be accelerated, to foster the removal of fossil fuels from the provision of heat, especially at low temperatures less than 300 degrees C. This paper looks at the results of two case studies for understanding the economics and potential for renewable thermal energy systems (RTES), particularly in hybrid configurations to provide industrial process heat (IPH). The first case study looks at heat pumps for district heating, and the second, the use of linear Fresnel collectors (LFCs) coupled with phase change material (PCM) thermal energy storage (TES) for direct steam generation (DSG). Using district heat as an input for the heat pump, three cases were run harvesting energy from ambient water (5 degrees C), sewage water (20 degrees C), and a solar collector (35 degrees C). Accounting for elevated costs of infrastructure for each heat source, the levelized cost of heat (LCOH) of the first case study ranged from $4-$10 per million British Thermal Units (MMBTU). For the second case study modeling LFCs with PCM and TES, the results show that a LCOH of $9-$15 per MMBTU is possible, depending on the direct normal irradiance.

concentrated solar thermal↗
DOE OSTI2017

Implementing a control application on an FPGA Platform - 356

Today, many I and C system suppliers propose FPGA-based I and C platforms. In that context, EDF has set up a project to assess the use of such platforms for safety-related systems and to enhance EDF's ability to manage this technology for the upgrade of existing I and C systems or for new constructions. One aim of the project is to assess the adequacy of these platforms for the development of control functions. This paper presents the development of a case study on a Radiy FPGA-based platform in a Hardware-In-the-Loop configuration, and the comparison with a Dymola Model-In-the-Loop and a Siemens T2000 S7 Hardware-In-the-Loop approach. The case study is a part of the level control of steam generators of French PWR NPP. The control flow diagram uses complex analog elementary blocks coded with fixed point format for the analog values. The use of IP cores of the FPGA supplier was avoided to enable a complete mastership of the VHDL code, to provide evidence of compliance with safety-related requirements, and to allow porting this code to another FPGA platform. The case study has highlighted the specificity of using a FPGA, and has shown that a complex control application in a FPGA-based PLC presents no major differences in precision and response time with a processor-based PLC. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Education and training in the field of nuclear instrumentation and measurement: CEA/INSTN (National Institute for Nuclear Sciences and Technologies) strategy to improve and develop new pedagogical tools and methods

Part of the French Alternative Energies and Atomic Energy Commission (CEA), the National Institute for Nuclear Science and Technology (INSTN) is a higher education institution whose mission is to provide students and professionals a high level of scientific and technological qualification in all disciplines related to nuclear energy applications. In this frame, INSTN carries out education and training (E and T) programs in nuclear instrumentation and radioprotection. Its strategy has always been to complete theoretical courses by training courses and laboratory works carried out on an extensive range of training tools that includes a large panel of nuclear instrumentation as well as software applications. Since its creation in 1956, the INSTN has conducted both education and vocational programs on ionizing radiation detection. An extensive range of techniques have commonly been used during practical works with students and employees of companies who need to get the knowledge and specialization in this field. Today, the INSTN is mainly equipped with usual detectors and electronics in large numbers in order to be able to accommodate up to 48 trainees at the same time in two classrooms, with only two trainees for one workstation in order to optimize their learning. In the field of the neutron detection systems, the INSTN has strongly developed its offer taking advantage of the use of research reactors, such as ISIS reactor (700 kW) at Saclay. The implementation of neutron detection systems specific to the courses offers a unique way of observing and analysing the signal coming from neutron detectors, as well as learning how to set the parameters of the detection system in real conditions. Providing the trainees with an extensive overview of each part of the neutron monitoring instrumentation apply to a nuclear reactor, hands-on measurements on the ISIS reactor play a major role in ensuring a practical and comprehensive understanding of the neutron detection system and their integration in the safety system of nuclear reactors. It also gives a solid background for the follow up and the development of the neutron detection systems. Another field of activity of the INSTN is the development of new teaching tools using software applications. We describe hereafter two applications that have recently been developed by the institute, O.S.I.R.I.S and DOSIMEX. The INSTN and the company OREKA have developed an innovative teaching tool named O.S.I.R.I.S. The tool is built on a virtual 3D environment in which users operate in a totally free way. The action is located in a pressurised water reactor in the environment of a steam generator building. The users are students or professionals who want to learn and train on concrete situations on how to protect the workers against radiation. Through the use of this serious game, users must: establish a predictive dose evaluation, implement the principles of radiation protection, perform the dose result of the operation, analyse the gap between the predictive collective dose and the collective dose achieved, and decide about the lessons to provide feedback. Taking into account the increasing constraints related to radiation protection, it is essential to use computer codes to evaluate the dose rates generated by sources of ionizing radiation. Many powerful codes exist but are often relatively complex to implement and are dedicated to experts. In addition, these codes are often 'black boxes' that does not allow the user to understand the underlying physics. The objective of DOSIMEX codes developed by the INSTN is to give, by the means of a VBA code with a simple interface, the ability to easily conduct a wide range of radiological exposure situations: calculation of gamma, beta, neutron equivalent dose rate, calculation of internal contamination and atmospheric transfer.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Simulation of Boiling Flow in a Helical Coil Using a Homogeneous Equilibrium Model with the Nek-2P CFD Code

Recent developments in the field of nuclear energy have lead to an increased interest in the design of helically coiled steam generators. In particular, their compact size, and high heat transfer rates make them ideal for small modular reactors. This has lead to a renewed effort to investigate the characteristics of fluid flow and heat transfer in and around helical coils both numerically and experimentally. The focus of this work is on the development of a consistent modeling, formulation, and simulation framework for boiling flows in helical coils using the highly scalable Nek-2P CFD code. As a first step, two phase flow is accounted for using a homogeneous equilibrium model. Results for both single and multiphase flows are compared to results obtained from Star-CCM+. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Performance Analysis of the Passive Residual Heat Removal System of an Integral Reactor

A demand for small modular reactors (SMRs) is gradually emerging worldwide, especially in Middle East Asia and Africa. SMART (System-integrated Modular Advanced ReacTor) developed by KAERI (Korea Atomic Energy Research Institute) is the first mover among several SMRs. It is designed to produce 100 MW of electricity, or 90 MW of electricity and 40,000 tons of desalinated water concurrently, which is sufficient for 100,000 residents. SMART is currently holding the initiative as the first licensed integral reactor in the world since 4 July 2012, when the standard design approval (SDA) was officially issued by the Nuclear safety and security commission (NSSC). After that, the SMART safety enhancement project was successfully conducted with intense support of the Government, resulting in an in-depth development of the most advanced safety features including the passive residual heat removal system (PRHRS). Recently, an agreement between KAERI and King Abdullah city for atomic and renewable energy (K.A. CARE) was officially signed for the pre-project engineering (PPE) in pursuit of the construction of SMART and developing human resources capability. Both parties are currently conducting a three-year preliminary study to review the feasibility of building SMART and to prepare for its commercialization. KAERI also takes responsibility to train R and D personnel chosen from Kingdom of Saudi Arabia. Sooner or later, SMART will open the global market of SMR and be the best competitive pioneer. Since the Fukushima accident, the passive cooling system of a nuclear reactor has received a great amount of attention. SMART is designed with passive systems in order to ensure the safety of the reactor. The PRHRS is one of the passive safety systems, which is activated after an accident to remove the residual heat from the core and the sensible heat of the reactor coolant system (RCS) through the steam generators (SGs) until the condition of the RCS reaches a safe shutdown condition. In the present study, the thermal performance of the PRHRS is predicted by using in-house code. Temperature of the RCS and the cooling rate of the RCS are traced with time to see if it is satisfactorily designed to meet the design requirements of the PRHRS. The dependency of the mass flow rate of the natural circulation on the filling ratio (FR) is also discussed. The FR is arbitrarily chosen from 0.22 to 0.50 and further parametric study is needed in the future.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

A Proposed Correlation for Critical Flow Rate of Steam/Water Flow

In the safety analysis of loss-of-coolant-accident (LOCA) and steam generator tube rupture (SGTR) scenarios of light water reactors, a modeling of a break is very important to predict the accident's result. For example, the design of a break simulator to simulate a LOCA in light water reactors requires accurate knowledge of the leakage flow through the break, whose shape can be assumed as an orifice or a nozzle according to the characteristics of the break shape. In the case of a small-break LOCA (SBLOCA) scenario, the critical flow mostly occurs by sub-cooled and saturated liquid water including a two-phase water flow at relatively high-pressure conditions. Among previous critical flow maps, Moody's maps are well known for their critical flow rate and pressure of a single component and two-phase mixture with respect to the upstream enthalpy and pressure. In Moody's model, a slip ratio defined by maximizing the critical flow rate with respect to the slip ratio was introduced, and thus its predictions tended to be too conservative. Later, Moody suggested other maps based on a homogeneous equilibrium model (HEM), and they were useful to predict a choked condition for a given upstream condition. However, their predictions of the critical flow rate seemed to show quite an underestimation compared to various test data. The author suggested more practicable critical flow maps for a critical mass flux and pressure of a steam-water flow with respect to the upstream stagnation conditions using an extended Henry-Fauske model. From a comparison with the selected test data, suggested maps showed conservative predictions within reasonable ranges. In this report, a new correlation for the critical flow rate was developed to simulate the suggested critical flow maps within a reasonable range of error. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Estimation of LOCA break size using cascaded support vector regression - 334

A failure of emergency core cooling system (ECCS) during a certain small break loss of coolant accident (SBLOCA) can cause a severe core uncovery and fuel failure. The safety injection systems (SISs) might not function properly in case of a SBLOCA, due to the slight change of pressure in the pipes. Early trend of the major parameters has to be observed and the accurate information has to be given to the nuclear power plant (NPP) operators by precisely identifying the break size to effectively manage accidents. In this study, the objective is to provide to operators the information on the break size in a short time by considering the accident situations such as hot-leg LOCA, cold-leg LOCA, steam generator tube rupture (SGTR). A cascaded support vector regression (CSVR) is used in order to estimate the break size. The simulation data set was obtained from the optimized power plant reactor 1000 (OPR1000) using modular accident analysis program (MAAP) code. And genetic algorithm (GA) was used to select the input variables of the CSVR model and optimize its parameters. As a result of this study, the CSVR model estimate very well the break size of LOCAs. If the operators can predict the break size in the LOCA, they can response quickly and properly to LOCA circumstances to prevent the core uncovery and fuel failure. Also, it will be possible to more efficiently manage beyond design basis accidents. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Predicting Rupture Pressure and Leak Rate of Alloy 600 U-bend Tubes with PWSCC and Comparing with Test Results (Technical Letter Report)

This report summarizes structural integrity work performed for the US NRC that predicts the structural integrity of steam generator (SG) tubes with flaws in the U-bend. The work covers model development and validation for U-bend tubes containing lab-induced primary water stress corrosion cracks (PWSCC). Limit load models for plastic collapse predicting ligament rupture pressure and leak rate of cracked U-bend tubes were developed and validated with pressure tests and finite element analyses (FEAs). Failure pressure and leak rate tests were conducted at room temperature on Alloy 600 U-bend tubes containing flaws located at various locations around the U-bend. Existing models for predicting structural integrity of straight tubes with well defined, rectangular, electro-discharge-machined (EDM) flaws were modified and used to predict the ligament rupture and unstable burst pressures as well as leak rates for the PWSCC flaws in the U-bend section.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

DTK C/Fortran Interface Development for NEAMS FSI Simulations

This report documents the development of DataTransferKit (DTK) C and Fortran interfaces for fluid-structure-interaction (FSI) simulations in NEAMS. In these simulations, the codes Nek5000 and Diablo are being coupled within the SHARP framework to study flow-induced vibration (FIV) in reactor steam generators. We will review the current Nek5000/Diablo coupling algorithm in SHARP and the current state of the solution transfer scheme used in this implementation. We will then present existing DTK algorithms which may be used instead to provide an improvement in both flexibility and scalability of the current SHARP implementation. We will show how these can be used within the current FSI scheme using a new set of interfaces to the algorithms developed by this work. These new interfaces currently expose the mesh-free solution transfer algorithms in DTK, a C++ library, and are written in C and Fortran to enable coupling of both Nek5000 and Diablo in their native Fortran language. They have been compiled and tested on Cooley, the test-bed machine for Mira at ALCF.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗