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NASA NTRSDate not supplied

Modeling Validation and Its Design Relevance for NTP Fuel Elements

The development of Nuclear Thermal Propulsion engines depends heavily on modeling and simulation of a number of disciplines—neutronic / fluid / thermal / structural are only a start. Further, rockets involve extreme conditions, particularly high material temperatures in the nuclear fuel. This paper considers validation of the underling physical models and codes and how their accuracy influences several design challenges. Several issues are identified, where models, material data, could enable design solutions. In particular, thermal and CTE mismatch stresses are a concern, and stress predictions would benefit from improved mechanical property data at high temperatures. Nuclear inter-element heating causes propellant flow maldistribution and performance risks, but a simple model provides insights. High temperature chemical diffusion should be modeled to understand fuel element mass loss.

Nuclear Propulsion↗
NASA NTRSDate not supplied

Modeling Validation and Its Design Relevance for NTP Fuel Elements

The development of Nuclear Thermal Propulsion engines depends heavily on modeling and simulation of a number of disciplines—neutronic / fluid / thermal / structural are only a start. Further, rockets involve extreme conditions, particularly high material temperatures in the nuclear fuel. This paper considers validation of the underling physical models and codes and how their accuracy influences several design challenges. Several issues are identified, where models, material data, could enable design solutions. In particular, thermal and CTE mismatch stresses are a concern, and stress predictions would benefit from improved mechanical property data at high temperatures. Nuclear inter-element heating causes propellant flow maldistribution and performance risks, but a simple model provides insights. High temperature chemical diffusion should be modeled to understand fuel element mass loss.

Nuclear Propulsion↗
DOE OSTI2020

A New Gold Mine? Achieving HVAC Energy Efficiency Through a System Metric

Washington State's Commercial Energy Code adopted a new energy metric called HVAC Total System Performance Ratio (TSPR) in 2019, a first in the codes world to regulate HVAC system efficiency. TSPR is a ratio of annual heating and cooling loads to the annual carbon emissions associated with the energy consumed by the HVAC system. TSPR provides a performance-based solution to evaluate and improve the overall HVAC design. The TSPR metric and its companion calculation tool were developed by Pacific Northwest National Laboratory (PNNL) with support from U.S. Department of Energy (DOE), Northwest Energy Efficiency Alliance (NEEA) and the City of Seattle. The new metric represents a significant shift in how HVAC design will meet code requirements. Utility programs can also leverage TSPR as a measure to determine energy savings and incentive amounts for HVAC retrofits. This paper describes the efforts by NEEA and its collaborators to prepare the market for TSPR adoption in code. This paper provides the pilot projects led by University of Washington Integrated Design Lab (IDL) to understand potential issues that could be faced by early adopters. This paper also covers how training and outreach provide engagement opportunities that can streamline code compliance, help address issues faced by early adopters and promote participation in utility programs. As Washington State works on the goals of 70% energy reduction and zero fossil-fuel greenhouse gas emission homes and buildings by the year 2031 , system level performance metrics will likely become increasingly more necessary and prevalent. This paper concludes that the HVAC TSPR requirement helps familiarize the HVAC industry with this approach and helps Washington achieve its long-term goals.

Liu, Bing↗
DOE OSTI2016

VERA Benchmarking Results for Watts Bar Nuclear Plant Unit 1 Cycles 1-12

The Consortium for Advanced Simulation of Light Water Reactors, the first US DOE Energy Innovation Hub, is developing and deploying a suite of directly coupled, high-fidelity physics methods known as VERA for performing engineering analyses of operating commercial nuclear power reactors. The centerpiece of VERA is the reactor core simulator, which employs the direct coupling of a 3D neutron transport code, a two-phase sub-channel thermal-hydraulic code, and a well-known isotopic depletion and decay code. Together these tools represent the most rigorous reactor simulator available today. Recently added capabilities include the depletion, shuffling, and discharge of fuel assemblies, enabling VERA to simulate multiple fuel cycles of operation. To demonstrate this capability, VERA has been used to simulate over 18 years of operation of TVA’s Watts Bar Nuclear Unit 1, and the results of these calculations have been used to quantify its performance in an industry-grade benchmarking activity. Comparisons are made with measured critical boron concentrations, control bank reactivity worths, isothermal temperature coefficients, and in-core neutron flux distributions. These results demonstrate that the advanced tools in VERA are capable for application to practical and common industrial problems.

Godfrey, Andrew T↗
DOE OSTI2024

Numerical simulation of involute-plate research reactor flow behavior using RANS, LES and DNS

This paper investigates the flow behavior of involute-plate research reactors by performing Reynolds-Averaged Navier Stokes simulation (RANS), Large Eddy Simulation (LES) and Direct Numerical Simulation (DNS) of the channel flow between fuel plates. By modeling turbulence with different numerical approaches, this study provides data with three levels of fidelity. For the RANS simulation, three widely used turbulence models, i.e., k-ε, k-ω, Reynolds Stress Turbulence model (RST) are applied by using the commercial CFD code STAR-CCM +. For LES and DNS, the open-source CFD code, Nek5000, is used given its outstanding scalability on High Performance Computer (HPC) and high-order technique. The results from RANS simulations are compared with that from LES and DNS for benchmarking. Both macroscale parameters and turbulence statistics, such as velocity magnitude, lateral velocity and turbulence kinetic energy, are presented and analyzed. The results from RANS simulation achieve good agreement with LES and DNS on velocity and turbulence kinetic energy prediction. The RST turbulence model predicts the most similar flow pattern of lateral velocity as compared to LES and DNS. The Lambda-2 (λ2) criterion with a reasonable threshold is used to demonstrate the instantaneous vortices distribution in the involute channel from both LES and DNS calculation. The DNS simulation captures more detailed turbulence especially near the corner, which explains the discrepancy between LES and DNS results near the corner. The normalized RMS error are defined and calculated to assess the performance of those turbulence models. The RST model captures the anisotropic feature of turbulence, which enable it to outperform other turbulence models for predicting the flow behavior in an involute channel. Although some discrepancies are found between LES and DNS results in the corner, the overall deviations between LES and DNS are found to be small. In conclusion, given that the computational cost of DNS calculation is an order of magnitude higher, using LES data for benchmarking RANS model is a cost-effective approach.

DNS↗
DOE OSTI2019

Micro Reactor Simulation Using the PROTEUS Suite in FY19

Two micro reactor cores were simulated using the MOC solver of PROTEUS: MegaPower for a heat pipe micro reactor and Holos-Quad for a high-temperature gas-cooled micro reactor. Verification of PROTEUS standalone was performed by simulating 2D and 3D core benchmark problems developed based on MegaPower and Holos-Quad. Simulation results indicated that PROTEUS successfully modelled 2D and 3D cores of the MegaPower and Holos-Quad micro reactors and PROTEUS eigenvalues agreed well with the SERPENT Monte Carlo solutions. For a heat pipe cooled micro reactor, multiphysics simulations were performed using PROTEUS / FLUENT / ANLHTP. The coupling system of the three codes were developed using Python-based drivers, which exchange the data (power, temperature, and heat rate) required for the coupling and control the boundary conditions of FLUENT and ANLHTP. The coupled system was qualitatively verified using a 3D unit assembly problem composed of six fuel rods and seven heat pipes, which was developed based on the MegaPower compositions. The steady-state problem was solved using FLUENT / ANLHTP with given asymmetric fuel powers, demonstrating the reasonable convergence of temperatures of 7 heat pipes. Transient simulations were also performed by making one out of 7 heat pipes fail. For the transient, the coupled system of PROTEUS / FLUENT / ANLHTP demonstrated reasonable changes of total power and heat pipe temperatures with time, accounting for temperature feedback effects. In this simulation, no effect of structural change was accounted for, which will be performed using ANSYS-mechanical in future along with further verification with larger or whole-core 3D problems.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

CASL COBRA-TF improvements for PWR DNB analysis

COBRA-TF (CTF) is a thermal hydraulic (T/H) subchannel code using either three-dimensional (3D) Cartesian or subchannel coordinate formulations for two-phase fluid flow and heat transfer solutions. CTF has been improved under the Consortium for Advanced Simulation of Light Water Reactors (CASL) program for Pressurized Water Reactor (PWR) applications, including software optimization, new closure models, pre- and post-processing and parallelization for modeling full reactor core T/H responses under normal operating and accident conditions. As a result of collaboration among CASL partners including the Westinghouse Electric Company, the Oak Ridge National Laboratory (ORNL), and the Sandia National Laboratories, additional modeling improvements were made to CTF specifically for PWR Departure from Nucleate Boiling (DNB) analysis, including a code option to evaluate fuel thermal margin in terms of DNB Ratio (DNBR) and an axial shape factor to account for effect of non-uniform axial power distribution on DNB. Multiple DNB correlations are now linked with CTF for different applications, including the Westinghouse proprietary WRB-1 correlation for fuel designs containing mixing vane grid spacers. The improved CTF code with the WRB-1 correlation (CTF/WRB-1) was validated using the DNB data from the PWR Subchannel Bundle Tests (PSBT). In addition to the comparison with the test data, the CTF/WRB-1 DNBR results and the associated local fluid conditions were compared to the results of the Westinghouse T/H design code, VIPRE-W, which is an enhanced version of the VIPRE-01 code originally developed by the Electric Power Research Institute (EPRI). The comparisons showed that CTF/WRB-1 DNBR predictions are in good agreement with the VIPRE-W results within the applicable range of the DNB correlation. A model sensitivity study was performed to confirm that the CTF void drift model had an insignificant effect on DNBR under the steam line break (SLB) low pressure condition. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Modeling of Heat-and-Mass Transfer in Fuel Assemblies in Liquid-Metal-Cooled Reactors with Partial Flow-Passage Blockage

The results of modeling the heat-and-mass transfer in the FA of liquid metal cooled reactors with partial flow-passage blockage are presented. The model employed for an isotropic porous body is described, and the obtained results are compared with experimental data and with direct numerical modeling performed with the CONV-3D code. It is shown that the porous-body model developed can be used to analyze threedimensional heat-and-mass transfer processes arising when the flow passage area of FA is partially blocked.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Impact of Time-Dependent Reactor and Sensor Physics on Core Power Synthesis (Rev.1)

Online synthesis of power distribution is critical in the operation and control of nuclear power reactors to ensure that the core is operating within safety margins and to provide essential knowledge associated with the burnup of the fuel. In light-water reactors, power synthesis is achieved by using some a priori knowledge of the state of the reactor core and updating based on the signals coming from in-core sensors—namely, self-powered neutron detectors (SPNDs). This report examines the effects of fuel burnup and sensor degradation on the ability to accurately synthesize the power distribution in a pressurized water reactor (PWR), considering the typical low-enriched uranium (LEU, 3%-5% enrichment) fuel cycle as well as the higher enrichment LEU+ (5%-8% enrichment) fuel cycle. Several modeling tools were used to simulate power synthesis based on the responses of SPNDs, with emitters made out of Rh or V. A representative PWR LEU core was modeled using the Polaris/Purdue Advanced Reactor Core Simulator (PARCS) approach. The Monte Carlo N-Particle Transport 6 (MCNP6) code was used as well to calculate response functions between different segments of fuel to individual SPNDs; this is a crucial parameter for power synthesis. The Oak Ridge Isotope GENeration (ORIGEN) package in the Standardized Computer Analyses for Licensing Evaluation (SCALE) code was used to model the time-dependent isotopic transmutation in the SPND emitters. All these data were fed into a custom code that enacted the point-based iterative method to simulate power synthesis. Developmental work was also performed on high-fidelity SPND models in the GEometry ANd Tracking 4 (Geant4) code, which enables higher-accuracy modeling of the current responses from SPNDs. In this work, five sets of time-dependent power synthesis test cases were conducted. In these test cases, systematic changes in the input conditions enabled an analysis of the effect of (1) slightly inaccurate a priori power distribution assumptions with respect to fuel burnup, (2) highly inaccurate a priori power distribution assumptions with respect to fuel burnup (such that burnup is not included in the a priori assumed distribution), and (3) differences between Rh and V SPNDs in terms of downstream consequences of the transmutation in the emitters and the extended nature of the LEU+ fuel cycle in comparison with LEU. The authors discovered that one may permissibly have slightly inaccurate a priori assumptions of the fuel burnup (such that the level of burnup may be slightly underapproximated or overapproximated by the accumulated burnup in approximately 9.3 full power days), but to not account for burnup at all in the a priori assumptions leads to severe levels of error, approaching 25% at maximum (for LEU). The authors also discovered that V SPNDs are extraordinarily robust in both the LEU and LEU+ fuel cycles considered in this modeling work, whereas Rh SPNDs undergo significant transmutation that can result in large errors in the synthesized power distribution.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Impact of Dutch nuclear fuel cycle scenarios on radionuclide inventory intended for disposal - 15201

One of the main subjects of Dutch five-year research program into geological disposal of radioactive waste, OPERA, is the post-closure radiological safety assessment. Reliable estimates of the radionuclide inventory and matrix composition are an important input for the safety assessment. The amounts and categories of waste foreseen to be disposed in 2130 the Netherlands have been compiled by the Dutch waste management organisation COVRA. Additionally, NRG has made a characterisation of the waste (radionuclides and matrix). This reference inventory is based on the Dutch base nuclear energy scenario of operating the Dutch single reactor, the Borssele NPP, until its presently scheduled shut-down in 2033. However, changes in the presently adopted nuclear fuel cycle strategy in the Netherlands may impact both the quantities of generated radioactive waste and its composition, and therefore the source term of the radiological safety assessment. To get a grip on this uncertainty a set of alternative future fuel cycle scenarios in the Netherlands, which are in compliance with scenarios formulated in the Dutch vision documents, has been considered and evaluated. The scenarios have been analysed with the computer code DANESS, 'Dynamic Analysis of Nuclear Energy System Strategies'. The result of the DANESS analyses concerns the amount and type of the radioactive waste to be finally disposed in a deep geological repository. This also includes the inventories of the individual radionuclides that are relevant for the post-closure safety assessment to be performed in a subsequent stage of the OPERA program. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Impact of Time Dependent Reactor and Sensor Physics on Core Power Synthesis

Online synthesis of the power distribution is critical in the operation and control of nuclear power reactors to ensure that the core is operating within safety margins, and to provide essential knowledge associated with the burnup of the fuel. In light water reactors (LWRs), power synthesis is achieved by using some a priori knowledge of the state of the reactor core and updating based on the signals coming from in-core sensors—namely, self-powered neutron detectors (SPNDs). This report aims to study the effects of fuel burnup and sensor degradation on the ability to accurately synthesize the power distribution in a LWR. Several modeling tools were used to simulate power synthesis based on the responses of SPNDs, with emitters made out of Rh or V. A representative pressurized water reactor low-enriched uranium (LEU) core was modeled using the Polaris/Purdue Advanced Reactor Core Simulator (PARCS) approach. The Monte Carlo N-Particle Transport 6 (MCNP6) code was used, as well, to calculate response functions between different segments of fuel to individual SPNDs; this is a crucial parameter for power synthesis. The Oak Ridge Isotope GENeration (ORIGEN) package in the Standardized Computer Analyses for Licensing Evaluation (SCALE) code was used to model the time-dependent isotopic transmutation in the SPND emitters. All these data were fed into a custom code that enacted the point-based iterative (PBI) method to simulate power synthesis. Developmental work was also performed on high-fidelity SPND models in the GEometry ANd Tracking 4 (Geant4) code, which enables higher-accuracy modeling of the current responses from SPNDs. In this work, five sets of time-dependent power synthesis test cases were conducted. In these test cases, systematic changes in the input conditions enabled an analysis of the effect of (1) slightly inaccurate a priori power distribution assumptions with respect to fuel burnup, (2) highly inaccurate a priori assumptions with respect to fuel burnup (such that burnup is not included in the a priori assumed distribution), and (3) differences between Rh and V SPNDs in terms of downstream consequences of the transmutation in the emitters. The authors discovered that one may permissibly have slightly inaccurate a priori assumptions of the fuel burnup (such that the level of burnup may be slightly under- or over-approximated by the accumulated burnup in approximately 9.3 full power days), but to not account for burnup at all in the a priori assumption leads to severe levels of error, approaching 25% at maximum. The authors also discovered that V SPNDs are extraordinarily robust in the low-enriched uranium fuel cycle considered in this modeling work, whereas Rh SPNDs undergo significant transmutation that can result in large errors in the synthesized power distribution.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2015

Combustion Characteristics in a Non-Premixed Cool-Flame Regime of n-Heptane in Microgravity

A series of distinct phenomena have recently been observed in single-fuel-droplet combustion tests performed on the International Space Station (ISS). This study attempts to simulate the observed flame behavior numerically using a gaseous n-heptane fuel source in zero gravity and a time-dependent axisymmetric (2D) code, which includes a detailed reaction mechanism (127 species and 1130 reactions), diffusive transport, and a radiation model (for CH4, CO, CO2, H2O, and soot). The calculated combustion characteristics depend strongly on the air velocity around the fuel source. In a near-quiescent air environment (< or = 2 mm/s), with a sufficiently large fuel injection velocity (1 cm/s), a growing spherical diffusion flame extinguishes at ≈1200 K due to radiative heat losses. This is typically followed by a transition to the low-temperature (cool-flame) regime with a reaction zone (at ≈700 K) in close proximity to the fuel source. The 'cool flame' regime is formed due to the negative temperature coefficient in the low-temperature chemistry. After a relatively long period (≈18 s) of the cool flame regime, a flash re-ignition occurs, associated with flame-edge propagation and subsequent extinction of the re-ignited flame. In a low-speed (≈3 mm/s) airstream (which simulates the slight droplet movement), the diffusion flame is enhanced upstream and experiences a local extinction downstream at ≈1200 K, followed by steady flame pulsations (≈0.4 Hz). At higher air velocities (4-10 mm/s), the locally extinguished flame becomes steady state. The present axisymmetric computational approach helps in revealing the non-premixed 'cool flame' structure and 2D flame-flow interactions observed in recent microgravity droplet combustion experiments.

diffusion flame↗
NASA NTRS2023

Subsonic Ultra Green Aircraft Research: Phase IV Final Report – Volume I Mach 0.80 Transonic Truss-Braced Wing High-Speed Design Report

This design report summarizes work performed by the Boeing Subsonic Ultra-Green Aircraft Research (SUGAR) team from December 2016 through June 2018. In Phase IV of the SUGAR program, Boeing is working to improve the productivity of a TTBW configuration by updating the Mach 0.745 vehicle concept to operate efficiently at Mach 0.80. Due to the aerodynamic and structural design challenges associated with adapting the design to a higher speed, design tasks have been broken into several subtasks. The information in this report (Volume I) documents the development of a Mach 0.80 vehicle outer mold line, maturation of TTBW structural concepts, and presents a preliminary high-liftsystem assessment. Results from these studies were integrated into an aircraft-level performance code that estimates the benefits of TTBW technology, both on its own and as a part of the SUGAR integrated technology suite. The aerodynamic performance results presented herein detail the high cruise efficiency of the TTBW concept operating at M=0.80. Results from the RANS code OVERFLOW indicate that the TTBW is aerodynamically well suited for operation at the design M=0.80 transonic cruise Mach number. Aircraft-level performance predictions show that the TTBW offers a 7.2% improvement in fuel burn (per seat) for an economic mission of 900nm, and a 9.0% improvement in fuel burn (per seat) for a 3500nm mission with full passenger payload, relative to an aspect ratio 13 conventional tube and wing configuration of equivalent technology. Integration of the TTBW performance benefits into the SUGAR technology suite shows a 55.9% reduction in fuel burn (for the economic mission) relative to the reference (SUGAR Free) aircraft, which is a consistently sized traditional cantilever wing, single-aisle transport typical of the 2008 fleet. This benefit does not include the performance improvement potential of hybrid-electric systems studied in earlier phases of the SUGAR program Structural concept development work completed in Phase IV has identified several strut attachment concepts that are worth additional development activity – an inboard strut attachment swivel joint concept (which enables the strut to carry axial loads while avoiding strut bending induced by wing twist during flight) significantly simplifies the construction of the main strut. An outboard offset strut attachment concept holds significant potential for further reductions in wing-strut interference drag. Based on work performed to date on Phase IV and the previous phases of the SUGAR program, the TTBW concept continues to remain a viable concept for significantly improving the performance of future aircraft. High-fidelity aerodynamic analysis of the Mach 0.80 TTBW aircraft has shown fuel burn improvement that has exceeded initial expectations despite the increase in cruise Mach number. Structural concept maturation has continued to refine the TTBW structure, and has found no obvious ‘show-stoppers’ to date. Therefore, the potential benefit of improved performance through the use of a very high aspect ratio strut-braced wing continues to warrant additional study

SUGAR↗
DOE OSTI2015

Performance assessment modeling of a generic SNF/HLW repository in salt with coupled thermal-hydrologic effects - 15423

This paper describes advances in performance assessment modeling of deep geologic repositories facilitated by a massively parallel, high-performance computing (HPC) environment. Our new Generic Disposal System Analysis (GDSA) Framework utilizes the massively parallel PFLOTRAN multi-physics code to simulate repository performance in the presence of coupled thermal-hydrologic-chemical processes, linked to DAKOTA, an HPC uncertainty sampling and propagation code that provides nested parallelism for multi-realization performance assessment and sensitivity analysis. This enhanced performance assessment (PA) modeling capability is demonstrated with deterministic and probabilistic simulations of a generic repository for SNF and HLW in bedded salt host rock, by comparing repository performance between a case with heat-generating waste ('thermal' case) and a case without heat generation ('isothermal' case). The simulation results provide preliminary insights into the effect of multi-physics processes and thermal-hydrologic coupling on the long-term behavior of a reference-case salt repository. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Serpent - Bison - THM Preliminary Multiphysics Modeling of a Nuclear Thermal Propulsion System Fuel Assembly

This work demonstrates the Monte Carlo neutronic and Thermo-Hydraulic coupling scheme using the Serpent code and MOOSE application Bison and Thermo Hydraulic Module. The coupling scheme is then applied to the reference BWX Technologies Nuclear Ther- mal Propulsion system at he fuel assembly level where it’s used to perform an analysis of the isothermal material coefficients and potential material reactivity worth. A method is developed to isolate which feedback effects should be considered for proceeding with reduced order deterministic neutronic modeling where branch off analysis must be con- ducted. The convergence behavior of the coupling scheme is demonstrated where it fol- lows the standard Picard iteration approach. Verification studies for the method of deduc- ing relevant feedback effects is also demonstrated.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

An Update on NiCE Support for BISON

The Nuclear Energy Advanced Modeling and Simulation program (NEAMS) from the Department of Energy s Office of Nuclear Energy has funded the development of a modeling and simulation workflow environment to support the various codes in its nuclear energy scientific computing toolkit. This NEAMS Integrated Computational Environment (NiCE) provides extensible tools and services that enable efficient code execution, input generation, pre-processing visualizations, and post-simulation data analysis and visualization for a large portion of the NEAMS Toolkit. A strong focus for the NiCE development team throughout FY 2015 has been support for the Multiphysics Object Oriented Simulation Environment (MOOSE) and the NEAMS nuclear fuel performance modeling application built on that environment, BISON. There is a strong desire in the program to enable and facilitate the use of BISON throughout nuclear energy research and industry. A primary result of this desire is the need for strong support for BISON in NiCE. This report will detail improvements to NiCE support for BISON. We will present a new and improved interface for interacting with BISON simulations in a variety of ways: (1) improved input model generation, (2) embedded mesh and solution data visualizations, and (3) local and remote BISON simulation launch. We will also show how NiCE has been extended to provide support for BISON code development.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

Improvements to neutron detection of Pu in solvent for an aqueous buffer tank, magnox reprocessing plant, Sellafield - 14216

The PP Active Feed Tank (PPAFT) is a buffer tank that is not safe by geometry and which is located between the Medium Active (MA) and Plutonium Purification (PP) cycles of the chemical separation process in the Magnox Reprocessing Plant. The PPAFT receives Pu nitrate in acid which has been conditioned ready for the extraction of Pu with TBP/OK (tri-butyl phosphate in kerosene) solvent in the PP Cycle. The criticality safety case for the PPAFT recognises that during both normal operations and potential maloperations solvent which is dissolved (TBP which is present in solution but can be released upon a change in acidity) and entrained (TBP/OK physically carried over but not in solution) will accumulate in the PPAFT. The only practicable route out for Pu in solvent in the PPAFT is the normal process route out from the base of the PPAFT to the PP cycle through intermediate vacuum lift and feed vessels. However, the dished end base of the PPAFT and the intermediate vacuum lift and feed vessels are of particularly challenging geometry. There are five neutron monitors placed around the PPAFT for the detection of Pu in solvent. The original design intent of the neutron detection system was that any single monitor should be able to reveal an accumulation of Pu in solvent before it would potentially be unsafe if fed downstream. However, over time there have been changes to the fuel feed and chemical process of the plant which have had the potential to affect the performance of the neutron monitoring system. The ability of the neutron monitoring system to function to its original intent was questioned. Work has recently been carried out to investigate the capabilities of the PPAFT neutron monitoring system for current plant conditions and to identify potential improvements while making best use of the existing neutron monitors. It was found that moving a neutron monitor from the side to the top of the PPAFT and adding alarms based on the ratio of signals from multiple neutron detectors would be effective in improving detection of Pu in solvent. This paper describes the challenges involved in detecting thin layers of Pu in solvent using neutron monitors, discusses the improvements made to the PPAFT neutron monitors and details the combination of real plant data and modelling with the shielding code MCBEND used to provide confidence in the enhanced neutron monitoring system. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Validation of the new code package APOLLO2.8 for accurate PWR neutronics calculations

This paper summarizes the Qualification work performed to demonstrate the accuracy of the new APOLLO2.S/SHEM-MOC package based on JEFF3.1.1 nuclear data file for the prediction of PWR neutronics parameters. This experimental validation is based on PWR mock-up critical experiments performed in the EOLE/MINERVE zero-power reactors and on P.I. Es on spent fuel assemblies from the French PWRs. The Calculation-Experiment comparison for the main design parameters is presented: reactivity of UOX and MOX lattices, depletion calculation and fuel inventory, reactivity loss with burnup, pin-by-pin power maps, Doppler coefficient, Moderator Temperature Coefficient, Void coefficient, UO{sub 2}-Gd{sub 2}O{sub 3} poisoning worth, Efficiency of Ag-In-Cd and B4C control rods, Reflector Saving for both standard 2-cm baffle and GEN3 advanced thick SS reflector. From this qualification process, calculation biases and associated uncertainties are derived. This code package APOLLO2.8 is already implemented in the ARCADIA new AREVA calculation chain for core physics and is currently under implementation in the future neutronics package of the French utility Electricite de France. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗