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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 217 records · Page 12
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Investigation of Benefits from U/TRU Recycle - Quantification and Comparison to U/Pu Recycle

The recently completed comprehensive evaluation and screening of nuclear fuel cycle options identified a number of potentially promising fuel cycles for R&D that offer what could be considered by decision-makers as having the potential for significant improvement compared to the current U.S. fuel cycle. The fuel cycles that consistently performed the best were recycle fuel cycles that used self-sustaining fast reactors operating with either U/Pu or U/TRU recycle fuel and also included options where the fast reactors provided fissile materials to support operation of thermal reactors. However, based on the evaluation criteria and metrics used in the study, there was no difference in benefit between recycle of U/Pu and U/TRU (where TRU is plutonium and the minor actinides) while there were differences in the challenges for developing and deploying such fuel cycles, with U/TRU recycle being more challenging. This observation prompted the question as to the desirability of pursuing R&D on U/TRU recycle given that there may not be an increase in benefit. As a result, activities have been pursued to further investigate the performance differences between U/Pu and U/TRU recycle based on considering issues beyond those used in the evaluation and screening study to identify, if possible, areas where there are significant benefits of U/TRU recycle compared to U/Pu recycle. These new considerations focused on several areas, but especially on the impact on disposal of the HLW, which in the case of U/Pu recycle contains all of the minor actinides along with fission products, while in the case of U/TRU recycle only contains the losses of minor actinides from the reprocessing and recycle fuel fabrication operations. This difference in content has several implications. One impact is on the time dependent decay heat which can affect handling and the use of space in a geologic repository. Another impact concerns the HLW form and volume, since presence of minor actinides may adversely affect the ability to reduce HLW volume. The short-term radioactivity and long-term radiotoxicity of the HLW is also affected, which may be of more or less importance depending on the specific geologic disposal environment. To study these potential effects, a range of waste forms and disposal environments were used in the analysis, documenting to what extent the recycle of minor actinides in addition to plutonium may offer further benefit. Another area of investigation concerned the recycle fuel, for the fast reactor and for the thermal reactors they may support. Information to date indicates that U/Pu fuel may be simpler to fabricate and has a much more extensive database than U/TRU fuel, one of the reasons for the increased challenge for developing and deploying a U/TRU fuel cycle, and also indicates that heterogeneous recycle of the minor actinides may be even more difficult as compared to homogeneous recycle. This information was reviewed and updated to reflect the most recent studies for the purpose of informing on all aspects of the differences between U/Pu and U/TRU recycle. The results of all of these investigations will be presented to provide information on the findings concerning the value of U/TRU recycle.

Fast reactor recycle↗
DOE OSTI2016

The performance of NSF in BWR operating conditions - 2016-0062

NSF is a zirconium alloy in the zirconium-tin-niobium-iron family. The nominal alloy content is 1 % tin, 1 % niobium, and 0.4 % iron. The initial targeted component for NSF in boiling water reactors (BWRs) is the fuel assembly channel. A BWR fuel channel encases the fuel bundle such that boiling is confined within the channel while the water remains solid outside the channel. The BWR fuel channel also defines the gap between the bundles where the cruciform control blade inserts and withdraws to control the criticality of the core. The key performance requirements of the BWR fuel channel are dimensional stability (including both shape and thickness) and adequate mechanical properties to withstand the applied stresses. The distortion mechanisms that affect dimensional stability are fluence gradient-induced bow, shadow corrosion-induced bow, and elastic and creep bulge. Although NSF is known to have improved irradiation growth characteristics compared with both Zircaloy-2 and Zircaloy-4 that translates to increased dimensional stability in fuel channels, the BWR corrosion performance, tensile properties, and irradiation creep performance are less well known. This paper reports on the physical metallurgy of NSF, oxide thickness and hydrogen content after in-reactor operation, mechanical properties (both irradiated and unirradiated), in-reactor creep response, and irradiation growth and bow of NSF channels. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2015

Irradiation of electronic components and circuits at the Portuguese Research Reactor: Lessons learned

The behavior of electronic components and circuits under radiation is a concern shared by the nuclear industry, the space community and the high-energy physics community. Standard commercial components are used as much as possible instead of radiation hard components, since they are easier to obtain and allow a significant reduction of costs. However, these standard components need to be tested in order to determine their radiation tolerance. The Portuguese Research Reactor (RPI) is a 1 MW pool-type reactor, operating since 1961. The irradiation of electronic components and circuits is one area where a 1 MW reactor can be competitive, since the fast neutron fluences required for testing are in most cases well below 10{sup 16} n/cm{sup 2}. A program was started in 1999 to test electronics components and circuits for the LHC facility at CERN, initially using a dedicated in-pool irradiation device and later a beam line with tailored neutron and gamma filters. Neutron filters are essential to reduce the intensity of the thermal neutron flux, which does not produce significant defects in electronic components but produces unwanted radiation from activation of contacts and packages of integrated circuits and also of the printed circuit boards. In irradiations performed within the line-of-sight of the core of a fission reactor there is simultaneous gamma radiation which complicates testing in some cases. Filters can be used to reduce its importance and separate testing with a pure gamma radiation source can contribute to clarify some irradiation results. Practice has shown the need to introduce several improvements to the procedures and facilities over the years. We will review improvements done in the following areas: - Optimization of neutron and gamma filters; - Dosimetry procedures in mixed neutron / gamma fields; - Determination of hardness parameter and 1 MeV-equivalent neutron fluence; - Temperature measurement and control during irradiation; - Follow-up of reactor power operational fluctuations; - Study of gamma radiation effects only. The fission neutron spectrum can be limitative for some of the tests, as most neutrons are in the 1-2 MeV energy range. Significant progress has been made lately in compact neutron generators using D-D and D-T fusion reactions, achieving higher neutron fluxes and longer lifetime than previously available. The advantages of using compact neutron generators for testing of electronic components and circuits will be also discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

ATRIUM{sup TM} 11 operating experience

AREVA NP developed the ATRIUM{sup TM}11 BWR fuel assembly to address the industry need for substantial improvements in fuel reliability and economic performance. The product development effort initially consisted of analytical design assessments and out-of-pile testing of prototype fuel assemblies and components. An integral component of the development project is a comprehensive in-pile demonstration program designed to fully exercise the new fuel assembly under a variety of operating conditions and to provide performance measurements of new design features. Fuel vendors, reactor operators, and regulators have successfully relied upon irradiation feedback derived from in-pile demonstration programs to safely introduce new features in reload quantities. The primary ATRIUM{sup TM} 11 in-pile demonstration program consists of 40 lead fuel assemblies (LFA) operating among five BWR reactors globally, covering the range of operational regimes and licensing environments necessary to confirm reload applications for the product. Since the initial irradiation in 2012, multiple pool-side inspection examinations have been performed. The inspection activities primarily focused on the new design features; including the spacer, fuel rod array, upper tie plate and fuel channel. These inspections have resulted in the collection of a substantial set of visual images, dimensional measurements, and performance feedback that are of general interest to the industry. Results from the inspections are compared to the database for existing products and to predetermined acceptance criteria. This paper provides details of the global in-pile demonstration program and the characteristics of the LFAs. Results of the inspection campaigns will be presented, including a summary of the technical assessment of specific new design features and the overall performance of the ATRIUM{sup TM} 11 product. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Results of the Phenix Dissymmetric Test Benchmark Exercise

In the context of the European H2020 project SESAME, a work package was dedicated to the validation of code coupling for the thermal-hydraulics of pool type reactors. To this end, CEA proposed a benchmark exercise of the so-called dissymmetric test performed at the end of life of the PHENIX reactor. The PHENIX reactor was an experimental reactor operated by CEA in France from 1974 until 2009. The knowledge and data acquired over the lifetime of this reactor are of paramount importance for the development of future sodium-cooled generation IV reactors. At the beginning of this test, the reactor is operated at nominal power (350 MWth) with 3 primary pumps and 4 intermediate heat exchangers. During this test, a pump trip occurs at the secondary side of one of the two loops (one loop provides coolant to two intermediate heat exchangers). The chain reaction is then stopped while the pumps remain operating at nominal speed. Dissymmetric flow patterns are then observed in the cold pool as hot sodium leaves the intermediate heat exchangers that are no longer cooled. The failed secondary pump is then restarted while the rotational speed of the other is reduced. Eventually, a stable homogeneous state is reached. This transient is challenging for both modelers and codes since this exercise tackles an industrial scale reactor featuring complex flows and interactions between different scales. The physical phenomena observed include thermal stratification, flow mixing and buoyancy effects. In order to accurately model these phenomena for a reasonable computational costs, the partners used System Thermal Hydraulics codes coupled with Computational Fluid Dynamics codes and one even included a subchannel code. In this paper, the PHENIX reactor and its behavior during the dissymmetric test are described. Then, a comparison of the different modeling options selected by the partners is offered. Finally, the results are compared to the experimental data and discussed and modelling recommendations are proposed.

code coupling↗
DOE OSTI2014

Confirmation of a realistic reactor model for BNCT dosimetry at the TRIGA Mainz

Purpose: In order to build up a reliable dose monitoring system for boron neutron capture therapy (BNCT) applications at the TRIGA reactor in Mainz, a computer model for the entire reactor was established, simulating the radiation field by means of the Monte Carlo method. The impact of different source definition techniques was compared and the model was validated by experimental fluence and dose determinations. Methods: The depletion calculation code ORIGEN2 was used to compute the burn-up and relevant material composition of each burned fuel element from the day of first reactor operation to its current core. The material composition of the current core was used in a MCNP5 model of the initial core developed earlier. To perform calculations for the region outside the reactor core, the model was expanded to include the thermal column and compared with the previously established ATTILA model. Subsequently, the computational model is simplified in order to reduce the calculation time. Both simulation models are validated by experiments with different setups using alanine dosimetry and gold activation measurements with two different types of phantoms. Results: The MCNP5 simulated neutron spectrum and source strength are found to be in good agreement with the previous ATTILA model whereas the photon production is much lower. Both MCNP5 simulation models predict all experimental dose values with an accuracy of about 5%. The simulations reveal that a Teflon environment favorably reduces the gamma dose component as compared to a polymethyl methacrylate phantom. Conclusions: A computer model for BNCT dosimetry was established, allowing the prediction of dosimetric quantities without further calibration and within a reasonable computation time for clinical applications. The good agreement between the MCNP5 simulations and experiments demonstrates that the ATTILA model overestimates the gamma dose contribution. The detailed model can be used for the planning of structural modifications in the thermal column irradiation channel or the use of different irradiation sites than the thermal column, e.g., the beam tubes.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Level monitoring system with pulsating sensor—Application to online level monitoring of dashpots in a fast breeder reactor

An innovative continuous type liquid level monitoring system constructed by using a new class of sensor, viz., pulsating sensor, is presented. This device is of industrial grade and it is exclusively used for level monitoring of any non conducting liquid. This instrument of unique design is suitable for high resolution online monitoring of oil level in dashpots of a sodium-cooled fast breeder reactor. The sensing probe is of capacitance type robust probe consisting of a number of rectangular mirror polished stainless steel (SS-304) plates separated with uniform gaps. The performance of this novel instrument has been thoroughly investigated. The precision, sensitivity, response time, and the lowest detection limit in measurement using this device are <0.01 mm, ∼100 Hz/mm, ∼1 s, and ∼0.03 mm, respectively. The influence of temperature on liquid level is studied and the temperature compensation is provided in the instrument. The instrument qualified all recommended tests, such as environmental, electromagnetic interference and electromagnetic compatibility, and seismic tests prior to its deployment in nuclear reactor. With the evolution of this level measurement approach, it is possible to provide dashpot oil level sensors in fast breeder reactor for the first time for continuous measurement of oil level in dashpots of Control and Safety Rod Drive Mechanism during reactor operation.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

Light Water Reactor Sustainability Program Materials Research Pathway Technical Program Plan

The Materials Research (MR) Pathway within the Light Water Reactor Sustainability (LWRS) Program is charged with performing the research and development (R&D) to develop the scientific basis for understanding and predicting long-term environmental degradation behavior of materials in nuclear reactors. Furthermore, it is essential to use the mechanistic understanding of degradation phenomena in materials to develop mitigation, repair, and new material alternatives for existing components. The work will provide data and methods to assess performance of systems, structures, and components (SSCs) essential to safe and sustained reactor operations. The R&D products developed from the LWRS Program will be used by utilities, industry groups, and regulators to inform operational and regulatory requirements for materials in reactor SSCs subjected to long-term operation conditions, providing key input to both regulators and industry. The intent of this research is to help in reducing the operating costs, which may be in the form of offset maintenance costs due to better predictive models for component lifetimes, improved analysis of materials through nondestructive evaluation, reduced costs for repairs, or extended performance of plants through the selection of improved replacement materials. To best achieve this, industry experience and guidance are important as is their role in coordinated or collaborative research projects. The objectives of this report are to describe the motivation and organization of the MR Pathway within the LWRS Program, provide details on the individual research tasks within the MR Pathway, describe the outcomes and deliverables of the MR Pathway, including recent technical highlights and progress, and list the requirements for performing the research.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2011

Cesium Isotope Ratios as Indicators of Nuclear Power Plant Operations

There are multiple paths by which radioactive cesium can reach the effluent from reactor operations. The radioactive 135Cs/137Cs ratios are controlled by these paths. In an effort to better understand the origin of this radiation, these 135Cs/137Cs ratios in effluents from three power reactor sites have been measured in offsite samples. These ratios are different from global fallout by up to six fold and as such cannot have a significant component from this source. A cesium ratio for a sample collected outside of the plant boundary provides integration over the operating life of the reactor. A sample collected inside the plant at any given time can be much different from this lifetime ratio. The measured cesium ratios vary significantly for the three reactors and indicate that the multiple paths have widely varying levels of contributions. There are too many ways these isotopes can fractionate to be useful for quantitative evaluations of operating parameters in an offsite sample, although it may be possible to obtain limited qualitative information for an onsite sample.

137Cs↗
DOE OSTI2021

Cyber Threat Assessment Methodology for Autonomous and Remote Operations for Advanced Reactors (Conference Presentation)

The next generation of Advanced Reactors include planned capabilities for both Autonomous (operating without human interaction for a set period-of-time) and Remote (operating with human interaction from a separate physical location) Operations. Existing Nuclear Reactor architectures include a set of safety and security constraints tightly coupled with personnel policies and procedures. As Advanced Reactors are fielded with these new Autonomous and Remote operational capabilities, the architectures and associated infrastructure services and components will perceivably expand the overall attack surface and risk calculations with regards to safe and secure operations. This paper is part of an FY21 work program focused on ensuring Advanced Reactor designs are informed with threat-based guidance on design and operation of Secure Architectures with a specific focus on the deployment of Autonomous Systems in support of Advanced Reactor Operations. The next phase of this research program is to complement produce a methodology for assessment of the cyber threat against these architectures as well as a catalogue of Use Cases to support the Advanced Reactor community in their implementation of Autonomous and Remote Operations.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2013

Preparation of the Second Shipment of Spent Nuclear Fuel from the Ustav Jaderneho Vyzkumu Rez (UJV Rez), a.s., Czech Republic to the Russian Federation for Reprocessing - 13478

After more than 50 years of operation of the LVR-15 research reactor operated by the UJV Rez, a. s. (formerly Nuclear Research Institute - NRI), a large amount of the spent nuclear fuel (SNF) of Russian origin has been accumulated. In 2005 UJV Rez, a. s. jointed the Russian Research Reactor Fuel Return (RRRFR) program under the United States (US) - Russian Global Threat Reduction Initiative (GTRI) and started the process of SNF shipment from the LVR-15 research reactor back to the Russian Federation (RF). In 2007 the first shipment of SNF was realized. In 2011, preparation of the second shipment of spent fuel from the Czech Republic started. The experience obtained from the first shipment will be widely used, but some differences must be taken into the account. The second shipment will be realized in 2013 and will conclude the return transport of all, both fresh and spent, high-enriched nuclear fuel from the Czech Republic to the Russian Federation. After the shipment is completed, there will be only low-enriched nuclear fuel on the territory of the Czech Republic, containing maximum of 20% of U-235, which is the conventionally recognized limit between the low- and high-enriched nuclear materials. The experience (technical, organizational, administrative, logistic) obtained from the each SNF shipment as from the Czech Republic as from other countries using the Russian type research reactors are evaluated and projected onto preparation of next shipment of high enriched nuclear fuel back to the Russian Federation. The results shown all shipments provided by the UJV Rez, a. s. in the frame of the GTRI Program have been performed successfully and safely. It is expected the experience and results will be applied to preparation and completing of the Chinese Miniature Neutron Source Reactors (MNSR) Spent Nuclear Fuel Repatriation in the near future. (authors)

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

An Evaluation of Standardized Canisters in the Waste Management System - 16317

Development of a standardized canister system represents an opportunity to develop an integrated approach to addressing storage, transportation, and disposal issues in the waste management system. However, regardless of timing and method, deployment of such a system would have the potential to cause significant system-wide impacts. This evaluation expands upon previous standardized-canister system evaluations and compares continued loading of dual-purpose canisters (i.e., the status quo) with loading of standardized canister systems in the near term, that is, before repository requirements are known and/or before operating reactors shut down. This evaluation quantitatively compares order-of-magnitude costs and logistics for different standardization scenarios with those of status quo scenarios, provides insight into quantifiable impacts of loading standardized canister systems in the near term, tests system-level analysis tools and associated input, and identifies scenarios for further analysis. This is a technical paper that does not take into account the contractual limitations under the Standard Contract (10 CFR Part 961). As work on assessing standardized canister system scenarios has progressed, new data for at-reactor and interim storage facility (ISF) operations have been collected and/or generated to provide more realism at the system level, and past scenarios were re-evaluated with this new information. In addition, new scenarios were developed and analyzed that focused on (1) bare fuel transported from reactors to an ISF in re-usable bare fuel casks and deferring deployment of a standardized canister system to the ISF, (2) alternative acceptance strategies based on conclusions from previous systems architecture studies, and (3) the impacts of having to accommodate spent fuel in differently sized canisters for different potential repository geologies. Results of this evaluation indicate that shifting the loading of standardized canisters from the reactor to the ISF does not result in a significant change in total system cost. Results also show that some alternative acceptance strategies could reduce the number of years that spent nuclear fuel (SNF) stays on-site at reactors for cases involving standardized canisters (consistent with previous results that looked at dual-purpose canisters). Finally, incorporating the new at-reactor loading data of small standardized canisters shows a significant system-wide cost reduction when compared to previous estimates of at-reactor loading of small canisters. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Minimization of Heat-Release Distribution Instability in Burning Nuclear Fuel

A method of solving the problem of minimizing the instability arising in the heat-release field as a result of fuel burnup and movement of the reactivity excess compensators during reactor operation. The computational particulars of the method and the results of solving the problem are discussed for a BN-1200 fast reactor (with two zones of different enrichment) where the motion of the compensators is modeled as a change in the volume fraction of the absorber.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Light Water Reactor Sustainability: Program Materials Research Pathway (FY22 Technical Program Plan)

The Materials Research (MR) Pathway within the Light Water Reactor Sustainability (LWRS) program is charged with performing R&D to develop the scientific basis for understanding and predicting the long-term environmental degradation behavior of materials in nuclear reactors. Furthermore, the mechanistic understanding of degradation phenomena in materials must be leveraged to develop mitigation and repair strategies and new material alternatives for existing components. This research will provide data and methods to assess the performance of systems, structures, and components essential to safe and sustained reactor operations. The R&D products developed from the LWRS program will be used by stakeholders—including utilities, industry groups, and regulators—to inform operational and regulatory requirements for materials in reactor systems, structures, and components subjected to long-term operation conditions, providing key inputs to regulators and industry. Therefore, the intent of this research is to provide options to reduce the operating costs, which may be in the form of offset maintenance costs due to better predictive models for component lifetimes, improved analyses of materials through nondestructive evaluation, reduced costs for repairs, or extended performance of plants through the selection of improved replacement materials. To provide the best options, industry experience and guidance are important because of their role in coordinated or collaborative research projects. The objectives of this report are to describe the motivation and organization of the MR Pathway within the LWRS program; provide details on the individual research tasks within the MR Pathway; describe the outcomes and deliverables of the MR Pathway, including recent technical highlights and progress; and describe the requirements for performing this critically important research.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Embedded acoustic sensing and monitoring techniques for small modular reactors

The goal of this project was to study the applicability of specific innovative instrumentation techniques for assessing parameters needed for safe small modular reactor operation and safeguarding of nuclear material. Small modular reactor core designs are currently being developed to provide energy more effectively and efficiently than in the past because they can be built as modules at fabrication sites and then transported to a power-producing facility. However, these modules and/or the final core will often be sealed and not accessible again until disposal. Instead of instruments that access the core directly, during and after operation, to monitor flux/dose and structural integrity, as current power plants use, new sensors need to be designed that can be built into the reactor initially to determine operating history, structural integrity through operation of the system, and nuclear material accountancy after shutdown and before disposition of the core. Embedded sensors already exist that can provide neutron flux, gamma dose, and temperatures, but techniques to expand upon these for assessing structural health and material inventory in the system over time need to be developed. Structural health assessments include the detection and imaging of cracks in the components that could eventually cause radioactive fission products to be released.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Fuel and cladding nano-technologies based solutions for long life heat-pipe based reactors

A novel nuclear reactor concept, unifying the fuel pipe with fuel tube functionality has been developed. The structure is a quasi-spherical modular reactor, designed for a very long life. The reactor module unifies the fuel tube with the heat pipe and a graphite beryllium reflector. It also uses a micro-hetero-structure that allows the fission products to be removed in the heat pipe flow and deposited in a getter area in the cold zone of the heat pipe, but outside the neutron flux. The reactor operates as a breed and burn reactor - it contains the fuel pipe with a variable enrichment, starting from the hot-end of the pipe, meant to assure the initial criticality, and reactor start-up followed by area with depleted uranium or thorium that get enriched during the consumption of the first part of the enriched uranium. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Model Calibration vs. Physics-guided Coverage Mapping: Case Study - Pressure Drop Calculations

This work deals with a fundamental question in model validation: given a body of available experiments and an envisaged domain of nuclear reactor operating conditions (referred to as the reactor application), how can one develop a quantitative approach that measures the portion of the prior uncertainties of the reactor application that is covered by the available experiments? Coverage here means that the uncertainties of the reactor application are originating from and behaving in exactly the same way as those observed at the experimental conditions. This approach is valuable as it provides a scientifically defendable criterion by which experimentally measured biases can be credibly extrapolated (i.e., mapped) to biases for the reactor applications. This work focuses on the application of a new methodology, referred to as physics-guided coverage mapping (PCM), to the estimation of hydraulic pressure losses in reactor pipes. PCM provides a generic, calibration-free approach to the mapping of experimental biases by relying on high fidelity physics simulation and information theoretic measure how much of the application domain is covered by the experiments. The idea is to employ the physics models to establish a joint probability distribution that relates the application and the available body of experiments. An information-theoretic measure, referred to as the mutual information, is used to calculate a quantitative metric that measures the inter-dependence between the application and the experiment. A virtual experiment is employed to demonstrate the application of PCM, wherein the pressure drop is calculated for both the experiment and the application using two different flow conditions, i.e., laminar and turbulent, and the experimental values are simulated. All parameters introducing uncertainties, such as geometry, friction factor formulas and their associated parameters are also incorporated. The results obtained with PCM are compared to a conventional calibration-based approach. This work show that the proposed approach can circumvent a number of challenges that face calibration-based methods, such as nonlinearity, ill-posedness and error compensation in high dimensional models with many parameters. This work focus primarily in this summary on parameter uncertainties and how they can be treated using PCM. In a separate summary, discuss performance of PCM under the presence of modeling uncertainties. This simple study has shown that calibration techniques for very simple situations cannot correctly scale the measurements from the experimental to the application domain. It is our view that is occurs primarily due the phenomenon of error compensation where on artificially changes the value of selected parameters to account for all sources of uncertainties. The situation gets more complicated when one calibrates multiple parameters where now the problem could potentially have an infinite number of ways by which the errors can be compensated. (authors)

42 ENGINEERING↗
DOE OSTI2014

Role of decommissioning plan and its progress for the PUSPATI TRIGA Reactor

Malaysian nuclear research reactor, the PUSPATI TRIGA Reactor, reached its first criticality in 1982, and since then, it has been serving for more than 30 years for training, radioisotope production and research purposes. Realizing the age and the need for its decommissioning sometime in the future, a ground basis of assessment and an elaborative project management need to be established, covering the entire process from termination of reactor operation to the establishment of final status, documented as the Decommissioning Plan. At international level, IAEA recognizes the absence of Decommissioning Plan as one of the factors hampering progress in decommissioning of nuclear facilities in the world. Throughout the years, IAEA has taken initiatives and drawn out projects in promoting progress in decommissioning programmes, like CIDER, DACCORD and R2D2P, for which Malaysia is participating in these projects. This paper highlights the concept of Decommissioning plan and its significances to the Agency. It will also address the progress, way forward and challenges faced in developing the Decommissioning Plan for the PUSPATI TRIGA Reactor. The efforts in the establishment of this plan helps to provide continual national contribution at the international level, as well as meeting the regulatory requirement, if need be. The existing license for the operation of PUSPATI TRIGA Reactor does not impose a requirement for a decommissioning plan; however, the renewal of license may call for a decommissioning plan to be submitted for approval in future.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗