scholarly journals NEUTRON EMISSION MEASUREMENTS OF PWR SPENT FUEL SEGMENTS AND PRELIMINARY VALIDATION OF DEPLETION CALCULATIONS

2021 ◽  
Vol 247 ◽  
pp. 10004
Author(s):  
G. Perret ◽  
D. Rochman ◽  
A. Vasiliev ◽  
H. Ferroukhi
Keyword(s):  
Fuel Cycle ◽  
Neutron Emission ◽  
Reactor Core ◽  
Minor Actinides ◽  
Emission Rates ◽  
Experimental Uncertainty ◽  
Spent Fuel ◽  
Preliminary Validation ◽  
Paul Scherrer ◽  
And Storage

Assessing neutron emission of LWR spent fuel is necessary for the back-end of the fuel cycle, such as the dimensioning of transport and storage casks of spent fuel. Although core and depletion codes can calculate the isotopic composition of the discharged fuel and therefore infer its neutron source, accurate measured neutron emission values remain rare mainly because of the difficulty to prepare, handle and characterize spent fuel. Measured neutron emission values are, however, extremely relevant to code validation, as neutrons emitted by LWR spent fuel mainly originates from spontaneous fissions of minor actinides (e.g., 242Cm, 244Cm and 252Cf) that are produced only after a large number of neutron captures in the reactor core. This paper reports on neutron emission measurements of selected LWR-PROTEUS spent fuel samples and their comparisons with a core and depletion calculation chains based on CASMO-5, SIMULATE-3 and the SNF codes. The measured LWR-PROTEUS samples are comprised of 11 samples irradiated in a Swiss PWR. The samples are UO2 or MOX and have discharge burn-ups ranging from 20 to 120 GWd/t. We measured the 40-cm long samples in a hot-cell of the Paul Scherrer Institut using a measurement station made of polyethylene and a BF3 detector. We repeated the measurements several times and in different conditions to ensure the accuracy and reproducibility of the results. We derived ratios of neutron rates emitted by the different samples and absolute neutron emission rates by comparison with a reference 252Cf source, which we re-calibrated for this exercise. The experimental uncertainty (1σ) on the absolute neutron emission varies from 3% to 4%. We compared a subset of the measured values to the calculation predictions and showed an agreement within less than 7% for all but one sample.

The Effect of MA Transmutation in the PWR on Fuel Cycle

2017 ◽  
Author(s):  
Haoyang Yu ◽  
Bin Liu ◽  
Wenxin Zhang ◽  
Jin Cai
Keyword(s):  
Fuel Cycle ◽  
Minor Actinide ◽  
Minor Actinides ◽  
Pressurized Water Reactor ◽  
Spent Fuel ◽  
Pressurized Water ◽  
Mcnp Program ◽  
And Control ◽  
Control Rods ◽  
Effective Multiplication

The minor actinides (MA) is important nuclides in the spent fuel which is bad for human ecological environment. Pressurized water reactor (PWR) is the main reactor type at commercial operation around world. It is important to find the appropriate loading patterns when introducing minor actinides to the PWR core. In this paper, we study the effect of MA transmutation in the PWR on fuel cycle. First, we use the MCNP program to simulate the model of PWR and the effective multiplication factor.Then,the MA is introduced into core in different ways and mass to simulate the effective multiplication factor. In conclusion,without considering chemical skim control and control rods, we change the thickness of the MA, until the keff closes to 1, We find that loading minor actinides to burnable poison rods for transmutation is an optimal minor actinide loading pattern.

scholarly journals Study of Minor Actinides Transmutation in PWR MOX Fuel

2017 ◽  
Author(s):  
Shengli Chen ◽  
Cenxi Yuan ◽  
Jingxia Wu ◽  
Yaolei Zou
Keyword(s):  
Nuclear Energy ◽  
Fuel Cycle ◽  
Minor Actinides ◽  
Pressurized Water Reactor ◽  
Spent Fuel ◽  
Pressurized Water ◽  
The Sustainable Development ◽  
Closed Nuclear Fuel Cycle ◽  
Mox Fuel ◽  
Management Method

The management of long-lived radionuclides in spent fuel is a key issue to achieve the closed nuclear fuel cycle and the sustainable development of nuclear energy. Partitioning-Transmutation is supposed to treat efficiently the long-lived radionuclides. Accordingly, the study of transmutation for long-lived Minor Actinides (MAs) is a significant work for the post-processing of spent fuel. In the present work, the transmutations in Pressurized Water Reactor (PWR) Mixed OXide (MOX) fuel are investigated through the Monte Carlo based code RMC. Two kinds of MAs are incorporated homogeneously into two initial concentrations MOX fuel assembly. The results indicate an overall nice efficiency of transmutation in both initial MOX concentrations, especially for two MAs primarily generated in the UOX fuel, 237Np and 241Am. In addition, the inclusion of 237Np has no large influence on other MAs, while the transmutation efficiency of 237Np is excellent. The transmutation of MAs in MOX fuel depletion is expected to be an efficient nuclear spent fuel management method.

scholarly journals An optimization methodology for heterogeneous minor actinides transmutation

2018 ◽  
Vol 4 ◽  
pp. 4
Author(s):  
Timothée Kooyman ◽  
Laurent Buiron ◽  
Gérald Rimpault
Keyword(s):  
Fuel Cycle ◽  
Volume Fraction ◽  
Negative Impact ◽  
Positive Impact ◽  
Optimal Solution ◽  
Minor Actinides ◽  
Spent Fuel ◽  
Closed Fuel Cycle ◽  
Decay Heat ◽  
The Core

In the case of a closed fuel cycle, minor actinides transmutation can lead to a strong reduction in spent fuel radiotoxicity and decay heat. In the heterogeneous approach, minor actinides are loaded in dedicated targets located at the core periphery so that long-lived minor actinides undergo fission and are turned in shorter-lived fission products. However, such targets require a specific design process due to high helium production in the fuel, high flux gradient at the core periphery and low power production. Additionally, the targets are generally manufactured with a high content in minor actinides in order to compensate for the low flux level at the core periphery. This leads to negative impacts on the fuel cycle in terms of neutron source and decay heat of the irradiated targets, which penalize their handling and reprocessing. In this paper, a simplified methodology for the design of targets is coupled with a method for the optimization of transmutation which takes into account both transmutation performances and fuel cycle impacts. The uncertainties and performances of this methodology are evaluated and shown to be sufficient to carry out scoping studies. An illustration is then made by considering the use of moderating material in the targets, which has a positive impact on the minor actinides consumption but a negative impact both on fuel cycle constraints (higher decay heat and neutron) and on assembly design (higher helium production and lower fuel volume fraction). It is shown that the use of moderating material is an optimal solution of the transmutation problem with regards to consumption and fuel cycle impacts, even when taking geometrical design considerations into account.

scholarly journals A concept for disposal of highly acidic spent nuclear fuel solutions at PSI

2019 ◽  
Vol 322 (3) ◽  
pp. 1857-1862 ◽  
Author(s):  
Mu Lin ◽  
Ivan Kajan ◽  
Dorothea Schumann ◽  
Andreas Türler
Keyword(s):  
Nuclear Fuel ◽  
Extraction Chromatography ◽  
Spent Nuclear Fuel ◽  
Paul Scherrer Institute ◽  
Minor Actinides ◽  
Flow Sheet ◽  
Spent Fuel ◽  
Dga Resin ◽  
Impregnated Resin ◽  
Paul Scherrer

Abstract During previous radioanalytical studies at Paul Scherrer Institute ca. 30 L of acidic waste containing spent nuclear fuel was produced, and now they need to be disposed A flow sheet for conditioning of these waste was designed and the extraction chromatography technique is evaluated. Suitable sorbents, such as AMP_PAN, TBP impregnated resin and DGA resin, were selected for the task of Cs-removal, extraction of U and Pu, and extraction of minor actinides and lanthanides, respectively. A pilot device will be built for preliminary tests with simulated solutions, and the facility will be built and evaluated with the real spent fuel solutions.

scholarly journals German Spent Nuclear Fuel Legacy: Characteristics and High-Level Waste Management Issues

2013 ◽  
Vol 2013 ◽  
pp. 1-11 ◽  
Author(s):  
A. Schwenk-Ferrero
Keyword(s):  
Nuclear Fuel ◽  
Power Plants ◽  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Fission Products ◽  
High Level Waste ◽  
Minor Actinides ◽  
Spent Fuel ◽  
Cycle Simulation ◽  
High Level

Germany is phasing-out the utilization of nuclear energy until 2022. Currently, nine light water reactors of originally nineteen are still connected to the grid. All power plants generate high-level nuclear waste like spent uranium or mixed uranium-plutonium dioxide fuel which has to be properly managed. Moreover, vitrified high-level waste containing minor actinides, fission products, and traces of plutonium reprocessing loses produced by reprocessing facilities has to be disposed of. In the paper, the assessments of German spent fuel legacy (heavy metal content) and the nuclide composition of this inventory have been done. The methodology used applies advanced nuclear fuel cycle simulation techniques in order to reproduce the operation of the German nuclear power plants from 1969 till 2022. NFCSim code developed by LANL was adopted for this purpose. It was estimated that ~10,300 tonnes of unreprocessed nuclear spent fuel will be generated until the shut-down of the ultimate German reactor. This inventory will contain ~131 tonnes of plutonium, ~21 tonnes of minor actinides, and 440 tonnes of fission products. Apart from this, ca.215 tonnes of vitrified HLW will be present. As fission products and transuranium elements remain radioactive from 104to 106years, the characteristics of spent fuel legacy over this period are estimated, and their impacts on decay storage and final repository are discussed.

Minor Actinides-Loaded FBR Core Concept Suitable for the Introductory Period

2008 ◽  
Author(s):  
Koji Fujimura ◽  
Akira Sasahira ◽  
Junichi Yamashita ◽  
Tetsuo Fukasawa ◽  
Kuniyoshi Hoshino
Keyword(s):  
Fuel Cycle ◽  
Feasibility Studies ◽  
Minor Actinides ◽  
Spent Fuel ◽  
Core Concept ◽  
Compact Type ◽  
The Core ◽  
Initial Loading ◽  
Start Up ◽  
Reactivity Coefficient

We proposed the “Flexible Fuel Cycle Initiative” (FFCI), which has flexibility for the uncertainties like the introduction speed of FBRs. On the other hand, during the FBR introduction period, Pu from LWR spent fuel is used for startup of FBRs. But the FBR core being loaded with Pu from LWR spent fuel has larger burnup reactivity due to its larger isotopic fraction of Pu-241 than the core being loaded with Pu from the FBR multi-recycling core. The increased burnup reactivity may reduce the cycle length of the FBR. In this paper, an FBR transitional core concept to handle the issues of the FBR introductory period was investigated. Core specifications are based on the compact type sodium-cooled MOX-fueled core designed in the Japanese FBR cycle feasibility studies, because the lower Pu inventory should be better for the FBR introductory period in view of its flexibility for the required reprocessing amount of LWR spent fuel to start up the FBR. The reference specifications are selected as follows. Output is 1500MWe and the average discharge fuel burnup is about 150GWd/t. Minor Actinides (MAs) recovered from LWR spent fuels which provide Pu to startup FBR are loaded to the initial loading fuels and exchanged fuels during some cycles until equilibrium. We set a kind of MA fraction rate of the initial loading fuel with 4 as the number of the fuel exchange batches. The average of the MA fraction of the initial loading fuel assumed is 3%, and the MA fraction of the exchange fuel is set as 5%. This 5% maximum of the MA fraction is based on the irradiation results of the experimental fast reactor Joyo. The core performance including burnup characteristics and reactivity coefficient were evaluated, and we confirmed that the transitional core from the initial loading until equilibrium cycle loaded Pu from LWR spent fuels could keep the resemble performance with the FBR multi-recycling core.

Transmutation Study of Minor Actinides in Mixed Oxide Fueled Typical Pressurized Water Reactor Assembly

2018 ◽  
Vol 4 (4) ◽  
Author(s):  
Shengli Chen ◽  
Cenxi Yuan
Keyword(s):  
Monte Carlo ◽  
Mixed Oxide ◽  
Fuel Cycle ◽  
Neutron Transport ◽  
Minor Actinides ◽  
Pressurized Water Reactor ◽  
Spent Fuel ◽  
Pressurized Water ◽  
Water Reactor ◽  
Mox Fuel

The management of long-lived radionuclides in spent fuel is a key issue to achieve the closed nuclear fuel cycle and the sustainable development of nuclear energy. The partitioning-transmutation method is supposed to efficiently treat the long-lived radionuclides. Accordingly, the transmutation of long-lived minor actinides (MAs) is significant for the postprocessing of spent fuel. In the present work, the transmutations in pressurized water reactor (PWR) mixed oxide (MOX) fuel are investigated through the Monte Carlo neutron transport method. Two types of MAs are homogeneously incorporated into MOX fuel assembly with different mixing ratios. In addition, two types of design of semihomogeneous loading of 237Np in MOX fuels are studied. The results indicate an overall nice efficiency of transmutation in PWR with MOX fuel, especially for 237Np and 241Am, which are primarily generated in the current uranium oxide fuel. In addition, the transmutation efficiency of 237Np is excellent, while its inclusion has no much influence on other MAs. The flattening of power and burnup are achieved by semihomogeneous loading of MAs. The uncertainties of Monte Carlo method are negligible, while those due to nuclear data change little the conclusions of the transmutation of MAs. The transmutation of MAs in MOX fuel is expected to be an efficient method for spent fuel management.

scholarly journals EVALUATION OF NEUTRON SHIELDING PERFORMANCE OF CD-SS 316L AS A CANDIDATE ALLOY FOR DRY CASK OF RESEARCH REACTOR SPENT FUEL

2021 ◽  
Vol 27 (2) ◽  
pp. 103
Author(s):  
Kuat Heriyanto ◽  
Usman Sudjadi ◽  
Jaka Rachmadetin ◽  
Yuli Purwanto ◽  
Pungky Ayu Artiani ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Research Reactor ◽  
Neutron Emission ◽  
Reactor Core ◽  
National Strategy ◽  
Spent Fuel ◽  
Neutron Shielding

EVALUATION OF NEUTRON SHIELDING PERFORMANCE OF CD-SS 316L AS A CANDIDATE ALLOY FOR DRY CASK OF RESEARCH REACTOR SPENT FUEL Development of dry casks is necessary to support the national strategy for management of spent fuels. One of the requirements for the dry cask is shielding performance for neutron emitted by the spent fuels to be stored in the dry cask. The objectives of this study are to determine the emitted neutrons by the spent fuel generated from GAS research reactor and to evaluate the neutron shielding performance of Cd-SS316L alloy as a candidate material to be used in dry cask for the spent fuels.  The former was carried out using Origen 2.1 software, while the latter using MCNP5. The result shows that the emitted neutrons by a spent fuel after 5 years discharged from GAS research reactor were 2.81×103 and 3.32×106 n/s for reactor core power of 15 and 30 MW, respectively. Addition of Cd improves the neutron shielding performance of SS 316L. The evaluation of neutron shielding performance of SS 316L with addition of Cd which is the candidate material for dry cask of the spent fuels from the GAS research reactor can be evaluated using Origen 2.1 software for neutron emission, while the neutron shielding performance was evaluated by the simulation using MNCP 5 software. This study shows the Cd-SS 316L alloy can be used for further study to develop the dry cask design for the GAS research reactor.Key words: Neutron shielding, cadmium, stainless steel, spent fuel.

The Sensitivity Simulation of Different Geometric Pretreatment in Method of Characteristics

2017 ◽  
Author(s):  
Tian XiaoRui ◽  
Zhou Tao ◽  
Li Zichao ◽  
Yu Tao
Keyword(s):  
Computational Efficiency ◽  
Method Of Characteristics ◽  
Azimuthal Angle ◽  
Reactor Core ◽  
Characteristic Line ◽  
Simulation Results ◽  
And Storage

In reactor core physics analysis,the research about the pre-processing of Method of Characteristic (MOC) including the generation and storage of characteristic line,the progress of calculation and the choosing of different quadrature set.In addition,doing some simulations,which is based on OpenMOC code and C5G7-MOX benchmark,about different parameters (including the track spacing,azimuthal angles and polar angles) and calculated its impacts on the computational efficiency and accuracy.the simulation results are as following:setting the track spacing as 0.1 cm or the azimuthal angle number as 4,the simulation results have better accuracy. Whether choosing the Leonard’s optimum quadrature set or the Tabuchi-Yamamoto quadrature set,the number of polar angles have tiny impact on accuracy.

scholarly journals The Need for Integrating the Back End of the Nuclear Fuel Cycle in the United States of America

MRS Advances ◽  
2018 ◽  
Vol 3 (19) ◽  
pp. 991-1003 ◽  
Author(s):  
Evaristo J. Bonano ◽  
Elena A. Kalinina ◽  
Peter N. Swift
Keyword(s):  
United States ◽  
Nuclear Fuel ◽  
United States Of America ◽  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Nuclear Fuel Cycle ◽  
The United States ◽  
Spent Fuel ◽  
Dry Storage ◽  
The Us

ABSTRACTCurrent practice for commercial spent nuclear fuel management in the United States of America (US) includes storage of spent fuel in both pools and dry storage cask systems at nuclear power plants. Most storage pools are filled to their operational capacity, and management of the approximately 2,200 metric tons of spent fuel newly discharged each year requires transferring older and cooler fuel from pools into dry storage. In the absence of a repository that can accept spent fuel for permanent disposal, projections indicate that the US will have approximately 134,000 metric tons of spent fuel in dry storage by mid-century when the last plants in the current reactor fleet are decommissioned. Current designs for storage systems rely on large dual-purpose (storage and transportation) canisters that are not optimized for disposal. Various options exist in the US for improving integration of management practices across the entire back end of the nuclear fuel cycle.

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