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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 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Modified versions of the fuel performance codes RTOP and RTOP-CA

RTOP code predicts thermal and mechanical behavior of LWR fuel rods and release of GFPs both in normal stationary and transient operation modes, and transient power ramps, specific to the reactivity-initiated accidents (RIA). Computational domains of the code include fuel column, zirconium cladding, gas under cladding, and fuel microstructure - UO{sub 2} grain and the development of gaseous porosity of different origin during irradiation. RTOP-CA code is based on the RTOP code. It simulates the behavior of a fuel rod with cladding rupture opening and RFPs release into the primary coolant circuit in normal operation modes. The paper presents the results of development and validation of engineering models, which take into account characteristics of the neutron spectrum during the fuel cycle and are integrated into modernized versions of the fuel performance codes RTOP and RTOP-CA. The paper focuses on the detailed description of two models. Four-group neutron-physical engineering model describes burn-out of gadolinium isotopes and calculates spatial power profile in a Gd-doped fuel rod pellet. The second engineering model presented is developed for fast estimation of fuel burnup in case of fuel rod cladding leakage accident. The comparison of the present calculations with the results of calculations by neutron codes shows that the model describes the heat rate distribution and the gadolinium burn-out with sufficient accuracy. The code also calculates the LHR of Gd-doped fuel rods given the heat rate of the adjacent UO{sub 2} fuel rods. This work allows determining the histories of LHR for the Gd-doped fuel rod from the average LHR of FA for various equilibrium WWER fuel cycles, power units and fuel enrichments. The presented model is implemented in the RTOP fuel performance code to calculate the thermomechanical behavior of the fuel rods. An engineering-physical model is developed for fast estimation of {sup 134}Cs/{sup 137}Cs activity ratio in fuel taking into account the parameters of the neutron spectrum during the fuel cycle. Implementation of the model enables to increase the reliability of correct fuel burnup determination in a failed fuel assembly (FA)based on the FERDS activity measurements during reactor operation. Results of the model calculations obtained for the FAs with a higher fuel enrichment and the NPP activity measurements have shown a higher {sup 134}Cs/{sup 137}Cs activity ratio compared to the values for the fuel with enrichment up to 3.6 wt. % in {sup 235}U. Currently, the engineering model is applicable within the range of UO{sub 2} enrichment from 3 to 5 wt. % in {sup 235}U, and Gd{sub 2}O{sub 3} enrichment from 5 to 8 wt. %.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Counterpart Test on IBLOCA of 13% Cold Leg Break in ATLAS Facility

In the analysis for Loss-of-Coolant Accident (LOCA) as a design-basis accident (DBA) of a pressurized water reactor (PWR), redefinition of the break size based on the risk-informed regulation (RIR) is being extensively investigated due to the potential safety benefits. The pipe integrity and break frequency has been found to depend on the pipe size, so that a complete rupture of a smaller pipe such as a pressurizer surge-line or an emergency core cooling system (ECCS) nozzle is more probable than a double-ended guillotine break (DEGB) of primary coolant circuit piping. The scenario with the break of these pipes can be regarded as intermediate break (IB) LOCA. With an aim of expanding code validation to address safety issues of the IBLOCA and developing an evaluation model (EM) for full spectrum LOCA, the database for the IBLOCA test in an integral effect test (IET) facility is essential. In particular, an experimental data from the counterpart test can contribute to enhance our understanding on the phenomena and resolve the scaling issues in the scaled-down test facilities. The objective of this study is to investigate thermal hydraulic phenomena during the IBLOCA by the counterpart test for the 13% cold leg break in ATLAS facility, which was the A.2 test in the OECDATLAS project. This integral effect test would contribute to evaluate the scaling effect of the ATLAS design and test condition.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

Analysis of fission gas release kinetics by on-line mass spectrometry

The release of fission gas (Xe and Kr) and helium out of nuclear fuel materials in normal operation of a nuclear power reactor can constitute a strong limitation of the fuel lifetime. Moreover, radioactive isotopes of Xe and Kr contribute significantly to the global radiological source term released in the primary coolant circuit in case of accidental situations accompanied by fuel rod loss of integrity. As a consequence, fission gas release investigation is of prime importance for the nuclear fuel cycle economy, and is the driven force of numerous R and D programs. In this domain, for solving current fuel behavior understanding issues, preparing the development of new fuels (e.g. for Gen IV power systems) and for improving the modeling prediction capability, there is a marked need for innovations in the instrumentation field, mainly for: . Quantification of very low fission gas concentrations, released from fuel sample and routed in sweeping lines. Monitoring of quick gas release variations by quantification of elementary release during a short period of time. Detection of a large range of atomic masses (e.g. H{sub 2}, HT, He, CO, CO{sub 2}, Ne, Ar, Kr, Xe), together with a performing separation of isotopes for Xe and Kr elements. Coupling measurement of stable and radioactive gas isotopes, by using in parallel mass spectrometry and gamma spectrometry techniques. To fulfill these challenging needs, a common strategy for analysis equipment implementation has been set up thanks to a recently launched collaboration between the CEA and the Univ. of Provence, with the technological support of the Liverpool Univ.. It aims at developing a chronological series of mass spectrometer devices based upon mass filter and 2D/3D ion traps with Fourier transform operating mode and having increasing levels of performances to match the previous challenges for out-of pile and in-pile experiments. The final objective is to install a high performance online mass spectrometer coupled to a gamma spectrometer in the fission product laboratory of the future Jules Horowitz Material Test Reactor. An intermediate step will consist of testing first equipment on an existing experimental facility in the LECA-STAR Hot Cell Laboratory of the CEA Cadarache. This paper presents the scientific and operational stakes linked to fission gas issues, resumes the current state of art for analyzing them in nuclear facilities, then presents the skills gathered through this collaboration to overcome technological bottlenecks. Finally it describes the implementation strategy in nuclear research facilities of the CEA Cadarache. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Thermal-hydraulic performance of a high-temperature zigzag-channel printed circuit heat exchanger - 18563

Advanced nuclear reactors, such as high-temperature gas-cooled reactors (HTGRs) from the Generation IV nuclear systems, are designed with the capability of delivering high-pressure, high-temperature helium to power conversion units for electricity generation and industrial plants for process heat applications. The efficiency of the electricity generation and process heat applications of HTGRs is critically dependent upon an intermediate heat exchanger (IHX), which is a key component in transferring thermal energy from the primary coolant to a secondary coolant. Printed circuit heat exchangers (PCHEs) are promising to be employed in HTGR designs due to their capability for high-temperature, high-pressure applications and compactness. In the current study, a laboratory-scale zigzag-channel PCHE was fabricated and its heat transfer and pressure drop characteristics were investigated experimentally in a high-temperature helium test facility (HTHF). A commercial computational fluid dynamics (CFD) code, STARCCM+, was used to simulate the thermal-hydraulic performance of the fabricated PCHE with a simplified geometry model. Comparisons showed some differences between experimental data available for the zigzag-channel PCHE and the numerical results. Local thermal-hydraulic performance analyses indicated that fully-developed flow conditions and periodic changes were not observed, which could be attributed to the temperature variations along the flow channels, resulting in large fluid property variations. In addition, the effects of sharp-edged or rounded zigzag channel geometry were numerically evaluated. The use of rounded corners at each bend of the channels is recommended since this method can reduce the pressure drop on both the hot and cold sides of the heat exchanger while maintaining similar heat transfer capabilities. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

The preliminary analysis on the steady-state and kinetic features of the molten salt pebble-bed reactor

A novel design concept of molten salt pebble-bed reactor with an ultra-simplified integral primary circuit called 'Nuclear Hot Spring' has been proposed, featured by horizontal coolant flow in a deep pool pebble-bed reactor, providing 'natural safety' features with natural circulation under full power operation and less expensive primary circuit arrangement. In this work, the steady-state physical properties of the equilibrium state of the molten salt pebble-bed reactor are calculated by using the VSOP code, and the steady-state thermo-hydraulic analysis is carried out based on the approximation of absolutely horizontal flow of the coolant through the core. A new concept of 2-dimensional, both axial and radial, multi-pass on-line fuelling scheme is presented. The result reveals that the radial multi-pass scheme provides more flattened power distribution and safer temperature distribution than the one-pass scheme. A parametric analysis is made corresponding to different pebble diameters, the key parameter of the core resistance and the temperature at the pebble center. It is verified that within a wide range of pebble diameters, the maximum pebble center temperatures are far below the safety limit of the fuel, and the core resistance is considerably less than the buoyant force, indicating that the natural circulation under full power operation is achievable and the ultra-simplified integral primary circuit without any pump is possible. For the kinetic properties, it is verified that the negative temperature coefficient is achieved in sufficient under-moderated condition through the preliminary analysis on the temperature coefficients of fuel, coolant and moderator. The requirement of reactivity compensation at the shutdown stages of the operation period is calculated for the further studies on the reactivity control. The molten salt pebble-bed reactor with horizontal coolant flow can provide enhanced safety and economical features. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Assessment on Transport of Radioactive Species from Nuclear Facilities - 17079

It is generally recognized that radioactive nuclides are most hazardous fission products that could be released in the event of reactor accidents. The reactor coolant system of a pressurized water reactor (PWR) nuclear plant exhibits radioactivity from various sources. These radiation sources consist of the fission products from fuel element and activated products from corrosion products. The complex infrastructure of the processes does not allow detailed measurement of activity until after the reactor is shutdown. All of the components that require dismantling will be radioactive which was caused by neutron activation during operation, contamination by corrosion products, contamination by damaged by fuel rods and leaks by equipment. Due mainly to the various operation modes and their aging, there are the increasing needs to carry out repair works that are not sufficiently provided for by the original design. Many difficulties are involved in these tasks, namely, high radiation doses, cramped areas, hard to reach components, coupled with stringent quality conditions. The need to simulate the behavior of corrosion products in a nuclear power plant environment has attracted many researchers' attention. It has performed the in-depth reviews of the various theories and physics associated with the corrosion products transportation, the laboratory loop testing, and the field measurements in operating plants. Before we start the decommissioning of nuclear power plants, workers, the public and the environment shall be properly protected from radiation hazards. For this purpose, a safety assessment should be carried out relating to the potential hazards that may occur during the decommissioning. The radiological hazards involved in the decommissioning activities also should be identified, and be taken into account in the safety assessment. The radiation conditions of the primary circuit are determined by the activity of corrosion and fission products that have accumulated on the internal surfaces of the components. Therefore the estimation of contamination on the components is very important and, may be the first step in the reactor operation and decommissioning of components. The predominant radionuclide in the PWR primary coolant is Co-60, although Co-58 contributes significantly to radioactivity level at the early stage of a reactor operation. Since Co-60 is produced from the in-core Co-58, which is the source of cobalt in a reactor coolant system, many researchers have investigated its transmutation in the core, its transport around the system, and its subsequent deposition on surfaces. The major parameter influencing the behavior of radioactivity, such as Co-60 in the coolant, is the corrosion product oxides. The objective of this work is to analyze the corrosion product activity transported in the primary circuit by using a multi-region model. This model, which is incorporated with the mass and activity transport between the dissolved corrosion products in the coolant and the surface, describes the specific activity of corrosion products in coolant and on the surfaces according to the operation modes as well as decommissioning stage. In order to evaluate the analysis of the corrosion products in an integral type reactor, a multi-region mass balance model is applied. The specific activities of the various species on the surface of core and steam generator throughout the effective full power hours, 15,000 hours, are simulated for the various modes of operations. As a result, the specific activities of corrosion products from the surface of primary circuit were shown as very low compared to those of other commercial nuclear power plants. In this simulated reactor, a significant reduction of specific activities from the activated corrosion products is obtained due to the introduction of titanium alloy to the steam generator tube and the avoidance of boric acid induced corrosion. From this result we conclude that we are able to predict the level of radioactive nuclides generated by corrosion products in a primary circuit according to the various modes of operations by using multi-region model. In this point of view it is concluded that this multi-region model can be used as an engineering tool for analysis of the corrosion products within the primary coolant for the nuclear power plants. It is also considered that it is possible to extend to a calculation tool for estimating the radiological hazards involved in the decommissioning activities in terms of safety assessment. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
NASA NTRS2007

Initial Testing of the Stainless Steel NaK-Cooled Circuit (SNaKC)

An actively pumped alkali metal flow circuit, designed and fabricated at the NASA Marshall Space Flight Center, is currently undergoing testing in the Early Flight Fission Test Facility (EFF-TF). Sodium potassium (NaK) was selected as the primary coolant. Basic circuit components include: simulated reactor core, NaK to gas heat exchanger, electromagnetic liquid metal pump, liquid metal flowmeter, load/drain reservoir, expansion reservoir, test section, and instrumentation. Operation of the circuit is based around the 37-pin partial-array core (pin and flow path dimensions are the same as those in a full core), designed to operate at 33 kWt. This presentation addresses the construction, fill and initial testing of the Stainless Steel NaK-Cooled Circuit (SNaKC).

Garber, Anne↗
NASA NTRS2007

Capabilities and Testing of the Fission Surface Power Primary Test Circuit (FSP-PTC)

An actively pumped alkali metal flow circuit, designed and fabricated at the NASA Marshall Space Flight Center, is currently undergoing testing in the Early Flight Fission Test Facility (EFF-TF). Sodium potassium (NaK), which was used in the SNAP-10A fission reactor, was selected as the primary coolant. Basic circuit components include: simulated reactor core, NaK to gas heat exchanger, electromagnetic (EM) liquid metal pump, liquid metal flowmeter, load/drain reservoir, expansion reservoir, test section, and instrumentation. Operation of the circuit is based around a 37-pin partial-array core (pin and flow path dimensions are the same as those in a full core), designed to operate at 33 kWt. NaK flow rates of greater than 1 kg/sec may be achieved, depending upon the power applied to the EM pump. The heat exchanger provides for the removal of thermal energy from the circuit, simulating the presence of an energy conversion system. The presence of the test section increases the versatility of the circuit. A second liquid metal pump, an energy conversion system, and highly instrumented thermal simulators are all being considered for inclusion within the test section. This paper summarizes the capabilities and ongoing testing of the Fission Surface Power Primary Test Circuit (FSP-PTC).

Garber, Anne E.↗
DOE OSTI2015

Characterization of spent filters from the water polishing system of IEA-R1 Reactor - 15451

The aim of this paper is to report studies performed to determine the activity of gamma emitters present in polypropylene cartridge filters from the polishing system of the IEA-R1 Reactor cooling-water primary circuit. The activities were calculated using the dose rates measured with hand-held detectors, after the ratios of the emission rates of photons were evaluated by gamma spectrometry, and the Point Kernel method, which correlates the activity of a source with the dose rates at various distances. The method described can be used to determine routinely the radioactive inventory of these filters, avoiding the necessity of destructive radiochemical analysis, or the necessity of calibrating the geometry of measurement. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

Multi-objective optimization of a PCHE-type intermediate heat exchanger using genetic algorithms

For the advanced nuclear reactor systems and power conversion units, it is important to optimize the design of intermediate heat exchangers (IHXs) which transfer heat from the primary loop to the secondary or intermediate loop. For a multi-objective optimization problem, a printed circuit heat exchanger-type IHX with zigzag and S-shaped fin channels is selected to be optimized in terms of both the total cost and the heat exchanger thermal effectiveness using genetic algorithms. A computer code is developed to search a set of optimized designs of the zigzag channel geometry. It is found that among the Pareto-optimal solutions, the total cost is approximately linearly related to the thermal effectiveness between 88 and 95% while it rises rapidly after the effectiveness exceeds 95%. An optimal flow channel diameter is also determined through the optimization. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

The Phenix ultimate natural convection test

The French sodium cooled fast reactor Phenix was shut down in 2009 after 35 years of operation. Before decommissioning, a final set of tests were performed by the CEA during 9 months. Several topics were involved such as thermal hydraulics, core physics and fuel behaviour. Among these ultimate experiments, two thermal hydraulic tests were performed: an asymmetrical test consisting in a trip of one secondary pump and a natural convection test in the primary circuit. Recognizing the unique opportunity offered by these Phenix ultimate tests, IAEA decided in 2007 to launch a Coordinated Research Project (CRP) devoted to benchmarking analyses with system codes on the Phenix natural convection test. One objective of the natural convection test in Phenix reactor is the assessment of the CATHARE system code for safety studies on future and advanced sodium cooled fast reactors. The aim of this paper is to describe this test, which was performed on June 22-23, 2009, and the associated benchmark specifications for the CRP work. The paper reminds briefly the Phenix reactor with the main physical parameters and the instrumentation used during the natural convection test. After that, the test scenario is described: - initial state at a power of 120 MWth, - test beginning resulting from a manual dry out of the two steam generators, - manual scram, - manual trip on the three primary pumps without back-up by pony motors, - setting and development of natural convection in the primary circuit, in a first phase without significant heat sink in the secondary circuits and in a second phase with significant heat sink in the secondary circuits, by opening the casing of steam generators to create an efficient heat sink, by air natural circulation in the steam generators casing. The benchmark case ends after this second phase, which corresponds to the experimental test duration of nearly 7 hours. The paper presents also the benchmark specifications data supplied by the CEA to all participants of this CRP in order to perform calculations (core, primary circuit, primary pumps, IHX, shutdown system, operating parameters, test scenario and real test conditions). Finally, main test results and analyses are presented including the evolution of the core and of the heat exchangers inlet and outlet temperatures, and some local temperature measurements. The natural convection has been easily set up in the pool type reactor Phenix with different boundary conditions at the secondary side, with or without heat sink. The data obtained during this unique test represent some very useful and precious results for the development of SFR in a wide range of thematic such as numerical methods dedicated to thermal-hydraulics safety analyses (system codes, CFD codes and coupling system and CFD codes) and instrumentation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Chooz-A steam generators characterization - 15205

EDF nuclear waste management requires a deep understanding of characterization, classification and waste sorting operations. In fact, French nuclear waste management defines several classes with specific management, treatment and storage facilities. Based on particular criteria, the more the radiological risk of the nuclear waste is important, the more its management will be complex and expensive. During the dismantling of the first French pressurized reactor Chooz-A, decontamination of the primary water circuits (not including the reactor vessel), the steam generators and the pressurizer have been carried out in order to reduce their activity levels. Thanks to these decontamination operations, and a specific characterization methodology, EDF was able to de-classify the 4 steam generators and store them in one piece at the ANDRA Very Low Level Activity disposal facilities instead of the Low and Intermediate Level activity one. This de-classification allowed EDF to avoid important cutting and packaging processes. Therefore, significant dose reductions for the operators and financial gain have been made. To characterize and declare Chooz-A SG activity, EDF-CIDEN uses a methodology defined by the French institute of atomic energy, CEA. The method is based on external gamma spectrometry measurements performed with NaI collimated detectors, associated with MERCURAD simulations providing the transfer functions for the detectors and activity sources. Internal measurements are carried out with a CZT (CdZnTe) probe inside the SG tubes to refine the 3D model. In fact, the primary side represents the main source of activity, and understanding its contamination distribution is important to reduce the model and calculation uncertainties. Measurements eventually provide SG {sup 60}Co global activity, from which the activities of other radionuclides of the spectrum are determined using scaling factors. The final activity declaration takes into account the standard deviation of the measurements in order to cover the uncertainties of the methodology. Thereby, the declaration is 'slightly over-estimative', enough to meet the waste storage facility acceptance criteria. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Strengthening of σ phase in a Fe20Cr9Ni cast austenite stainless steel

The strengthening mechanism of σ phase in a Fe20Cr9Ni cast austenite stainless steel used for primary coolant pipes of nuclear power plants has been investigated. The yield and ultimate tensile strengths of aged specimens increased comparing with those of the unaged ones. It was found that the increase of strengths is due to the hard and brittle (σ + γ{sub 2}) structure which decomposed from α phase in the steel. Fracture surfaces of specimens after in situ tensile test showed that the inhibition of (σ + γ{sub 2}) structure on the dislocation movements was more significant than ferrite although cracks started predominately at σ/γ{sub 2} interfaces. The (σ + γ{sub 2}) structure behaves like a fiber reinforced composite material. - Highlights: • The strengthening mechanism of σ phase in a Fe20Cr9Ni CASS is investigated. • The yield and ultimate tensile strengths increase with increasing of σ phase. • The increase of strengths is due to hard and brittle (σ + γ{sub 2}) structure. • The (σ + γ{sub 2}) structure in CASS behaves like a fibre reinforced composite material. • The σ/γ{sub 2} and α/σ/γ{sub 2} boundaries hinder the movement of dislocation.

36 MATERIALS SCIENCE↗
DOE OSTI2010

The Concentration Of Tritium In Urine And Internal Radiation Dose Estimation Of PTNBR Radiation Workers

The operation of Triga 2000 reactor in Nuclear Technology Center for Materials and Radiometry (PTNBR BATAN) normally produce tritium radionuclide which is the activation product of deuterium atom in reactor primary cooling water. According to previous monitoring, tritium was detected with the concentration of 8.236{+-}0.677 kBq/L and 1.704{+-}0.046 Bq/L in the primary cooling water and in reactor hall air, respectively. The tritium in reactor hall air chronically can be inhaled by the workers. In this research, tritium content in radiation workers' urine was determined to estimate the internal radiation doses received by the workers. About 50-100 mL of urine samples were collected from 48 PTNBR workers that is classified as 24 radiation workers and 24 administration staffs as a control. Urine samples of 25 mL were then prepared by active charcoal and KMnO{sub 4} addition and followed with complete distillation. The 2 mL of distillate was added with 13 mL scintillator, shaked vigorously and remained in cool and dark condition for about 24 hours. The tritium in the samples was then measured using liquid scintillation counter (LSC) for 1 hour. From the measurement results it was obtained that the tritium concentration in the urine of radiation workers were in the range of not detected and 5.191 Bq/mL, whereas in the administration staffs the concentration were between not detected and 4.607 Bq/mL. Internally radiation doses were calculated using the tritium concentration data, and it was found the averages about 0.602 {mu}Sv/year and 0.532 {mu}Sv/year for radiation workers and administration staffs, respectively. The doses received by the workers were lower than that of the permissible doses from tritium, i.e. 40 {mu}Sv/year.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Accident analysis of heavy water cooled thorium breeder reactor

Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The power reactor has a peak value before reactor has new balance condition. The analysis showed that temperatures of fuel and claddings during accident are still below limitations which are in secure condition.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Analysis of N-16 concentration in primary cooling system of AP1000 power reactor

Nitrogen-16 (N-16) is one of the radiation safety parameter on the primary reactor system. The activation product, N-16, is the predominant contributor to the activity in the reactor coolant system during reactor operation. N-16 is activation product derived from activation of O-16 with fast neutron based on {sup 16}O(n,p){sup 16}N reaction. Thus study is needed and it performs to determine N-16 concentration in reactor coolant (primary coolant) in supporting radiation safety. One of the way is using analytical methode based on activation and redecay princip to obtain N-16 concentration. The analysis was performed on the configuration basis and operational of Westinghouse AP1000 power reactor in several monitoring points at coolant reactor system. The results of the calculation of N-16 concentration at the core outlet, reactor vessel outlet, pressurizer line, inlet and outlet of steam generators, primary pumps, reactor vessels inlet and core inlet are: 281, 257, 255, 250, 145, 142, 129 and 112 µCi/gram respectively. The results of analysis compared with AP1000 design control document as standard values. The verification showed very high accuracy comparation between analytical results and standard values.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗