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At least 1,063 records · Page 59
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12

Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF{sub 2}-ThF{sub 4}-{sup 233}UF{sub 4} respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 data library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

A Summary of the Department of Energy's Advanced Demonstration and Test Reactor Options Study

An assessment of advanced reactor technology options was conducted to provide a sound comparative technical context for future decisions by the Department of Energy (DOE) concerning these technologies. Strategic objectives were established that span a wide variety of important missions and advanced reactor technology needs were identified based on recent Department of Energy and international studies. A broad team of stakeholders from industry, academia and government was assembled to develop a comprehensive set of goals, criteria and metrics to evaluate advanced irradiation test and demonstration reactor concepts. Point designs of a select number of concepts were commissioned to provide a deeper technical basis for evaluation. The technology options were compared on the bases of technical readiness and the ability to meet the different strategic objectives. Using the study’s evaluation criteria and metrics, an independent group of exerts from industry, universities and national labs scored each of the point designs. Pathways to deployment for concepts of varying technical maturity were estimated for the different demonstration systems with regard to cost, schedule and possible licensing approaches. Furthermore, this study also presents the tradeoffs that exist among the different irradiation test reactor options in terms of the ability to conduct irradiations in support of advanced reactor research and development (R&D) and to serve potential secondary missions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

10 CFR 830 Major Modification Determination for Advanced Test Reactor RDAS and LPCIS Replacement

The replacement of the ATR Control Complex's obsolete computer based Reactor Data Acquisition System (RDAS) and its safety-related Lobe Power Calculation and Indication System (LPCIS) software application is vitally important to ensure the ATR remains available to support this national mission. The RDAS supports safe operation of the reactor by providing 'real-time' plant status information (indications and alarms) for use by the reactor operators via the Console Display System (CDS). The RDAS is a computer support system that acquires analog and digital information from various reactor and reactor support systems. The RDAS information is used to display quadrant and lobe powers via a display interface more user friendly than that provided by the recorders and the Control Room upright panels. RDAS provides input to the Nuclear Engineering ATR Surveillance Data System (ASUDAS) for fuel burn-up analysis and the production of cycle data for experiment sponsors and the generation of the Core Safety Assurance Package (CSAP). RDAS also archives and provides for retrieval of historical plant data which may be used for event reconstruction, data analysis, training and safety analysis. The RDAS, LPCIS and ASUDAS need to be replaced with state-of-the-art technology in order to eliminate problems of aged computer systems, and difficulty in obtaining software upgrades, spare parts, and technical support. The major modification criteria evaluation of the project design did not lead to the conclusion that the project is a major modification. The negative major modification determination is driven by the fact that the project requires a one-for-one equivalent replacement of existing systems that protects and maintains functional and operational requirements as credited in the safety basis.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Use Cases and Model Development of Thermal Storage Coupling for Advanced Nuclear Reactors

This report discusses the different options for coupling thermal energy storage (TES) systems to advanced nuclear power plants (A-NPPs) in order to enable flexible and hybrid plant operation. An advanced light-water reactor (ALWR) and a high-temperature gas-cooled reactor (HTGR) were selected as the initial use cases for demonstrating a thermally balanced energy storage coupling design for thermal power extraction. Cost functions for the A-LWR were derived from the fully balanced models that were developed based on three different coupling options with three different thermal energy bypass ratios. For the next steps, cost functions for the HTGR will also be derived, and additional nuclear reactors (e.g., a liquid-cooled fast reactor [LFR] or molten-salt reactor [MSR]) will be evaluated for coupling with TES in similar fashion, including the evaluation of their steady-state condition models and cost functions. The models presented herein showcase several design considerations, focusing on optimal deployment methodologies for achieving steady-state operation with minimum disruption to the nuclear power generation cycle. This report presents the results of steady state models developed using Aspen HYSYS®, wherein the thermal energy bypass for an NPP-TES coupling was varied up to 50%. The various components were sized using Aspen Process Economic Analyzer (APEA) and Aspen Exchanger Design and Rating (EDR), when applicable. Cost functions from these models were developed using the latest publicly available data obtained from APEA V11. The current steady-state models and cost functions provide a baseline for additional work focusing on dynamic operation and process optimization by using Idaho National Laboratory (INL)’s Framework for Optimization of Resources and Economics (FORCE) tools to evaluate the technoeconomic viability and transient operations of TES-coupled A-NPPs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Validation of FSP Reactor Design with Sensitivity Studies of Beryllium-Reflected Critical Assemblies

The baseline design for space nuclear power is a fission surface power (FSP) system: sodium-potassium (NaK) cooled, fast spectrum reactor with highly-enriched-uranium (HEU)-O 2 fuel, stainless steel (SS) cladding, and beryllium reflectors with B4C control drums. Previous studies were performed to evaluate modeling capabilities and quantify uncertainties and biases associated with analysis methods and nuclear data. Comparison of Zero Power Plutonium Reactor (ZPPR)-20 benchmark experiments with the FSP design indicated that further reduction of the total design model uncertainty requires the reduction in uncertainties pertaining to beryllium and uranium cross-section data. Further comparison with three beryllium-reflected HEU-metal benchmark experiments performed at the Oak Ridge Critical Experiments Facility (ORCEF) concluded the requirement that experimental validation data have similar cross section sensitivities to those found in the FSP design. A series of critical experiments was performed at ORCEF in the 1960s to support the Medium Power Reactor Experiment (MPRE) space reactor design. The small, compact critical assembly (SCCA) experiments were graphite- or beryllium-reflected assemblies of SS-clad, HEU-O 2 fuel on a vertical lift machine. All five configurations were evaluated as benchmarks. Two of the five configurations were beryllium reflected, and further evaluated using the sensitivity and uncertainty analysis capabilities of SCALE 6.1. Validation of the example FSP design model was successful in reducing the primary uncertainty constituent, the Be(n,n) reaction, from 0.28 %$\delta$k/k to 0.0004 %$\delta$k/k. Further assessment of additional reactor physics measurements performed on the SCCA experiments may serve to further validate FSP design and operation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Preliminary safety calculations to improve the design of Molten Salt Fast Reactor

Molten salt reactors are liquid fuel reactors so that they are flexible in operation but very different in the safety approach from solid fuel reactors. This study bears on the specific concept named Molten Salt Fast Reactor (MSFR). Since this new nuclear technology is in development, safety is an essential point to be considered all along the R and D studies. This paper presents the first step of the safety approach: the systematic description of the MSFR, limited here to the main systems surrounding the core. This systematic description is the basis on which we will be able to devise accidental scenarios. Thanks to the negative reactivity feedback coefficient, most accidental scenarios lead to reactor shut down. Because of the decay heat generated in the fuel salt, it must be cooled. After the description of the tools developed to calculate the residual heat, the different contributions are discussed in this study. The decay heat of fission products in the MSFR is evaluated to be low (3% of nominal power), mainly due to the reprocessing that transfers the fission products to the gas reprocessing unit. As a result, the contribution of the actinides is significant (0.5% of nominal power). The unprotected loss of heat sink transients are studied in this paper. It appears that slow transients are favorable (> 1 min) to minimize the temperature increase of the fuel salt. This work will be the basis of further safety studies as well as an essential parameter for the design of the draining system. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Fast reactor 3D core and burnup analysis using VESTA

Burnup analyses using the VESTA code have been performed on a MOX-fuelled fast reactor model as specified by an IAEA computational benchmark. VESTA is a relatively new code that has been used for burnup credit calculations and thermal reactor models, but not typically for fast reactor applications. The detailed input and results of the IAEA benchmark provides an opportunity to gauge the use of VESTA in a fast reactor application. VESTA employs an ultra-fine multi-group binning approach that accelerates Monte Carlo burnup calculations. Using VESTA to compute the end of cycle (EOC) power fractions by enrichment zone showed agreement with the published values within 5%. When comparing the ultra-fine multi-group binning approach to the tally-based approach, EOC isotopic masses also agree within 5%. Using the ultra-fine multi-group binning approach, we obtain a wall-time speedup factor of 35 when compared to the tally-based approach for computing a k{sub eff} eigenvalue with burnup problem. The authors conclude the use of VESTA's ultra-fine multi-group binning approach with Monte Carlo transport performs accurate depletion calculations for this fast reactor benchmark. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

The technical and economic impact of minor actinide transmutation in a sodium fast reactor

Within the frame work of the French National Act of June 28, 2006 pertaining to the management of high activity, long-lived radioactive waste, one of the proposed processes consists in transmuting the Minor Actinides (MA) in the radial blankets of a Sodium Fast Reactor (SFR). With this option, we may assess the additional cost of the reactor by comparing two SFR designs, one with no Minor Actinides, and the other involving their transmutation. To perform this exercise, we define a reference design called SFRref, of 1500 MWe that is considered to be representative of the Reactor System. The SFRref mainly features a pool architecture with three pumps, six loops with one steam generator per loop. The reference core is the V2B core that was defined by the CEA a few years ago for the Reactor System. This architecture is designed to meet current safety requirements. In the case of transmutation, for this exercise we consider that the fertile blanket is replaced by two rows of assemblies having either 20% of Minor Actinides or 20% of Americium. The assessment work is performed in two phases. - The first consists in identifying and quantifying the technical differences between the two designs: the reference design without Minor Actinides and the design with Minor Actinides. The main differences are located in the reactor vessel, in the fuel handling system and in the intermediate storage area for spent fuel. An assessment of the availability is also performed so that the impact of the transmutation can be known. - The second consists in making an economic appraisal of the two designs. This work is performed using the CEA's SEMER code. The economic results are shown in relative values. For a transmutation of 20% of MA in the assemblies (S/As) and a hypothesis of 4 kW allowable for the washing device, there is a large external storage demanding a very long cooling time of the S/As. In this case, the economic impact may reach 5% on the capital part of the Levelized Unit Electricity Cost (LUEC). A diminished concentration at 10% of MA, reduces the size of the external storage and the cooling time of the assemblies becomes compatible with the management of the irradiated fuel. Even with a low allowable power for the washing device, the economic impact on the capital cost is less than 2.5%. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Testing of a Transport Cask for Research Reactor Spent Fuel - 13003

Since the beginning of the last decade three Latin American countries that operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the TRIGA and MTR reactors operated in the region. A main drive in this initiative, sponsored by the International Atomic Energy Agency, is the fact that no definite solution regarding the back end of the research reactor fuel cycle has been taken by any of the participating country. However, any long-term solution - either disposition in a repository or storage away from reactor - will involve at some stage the transportation of the spent fuel through public roads. Therefore, a licensed cask that provides adequate shielding, assurance of subcriticality, and conformance to internationally accepted safety, security and safeguards regimes is considered a strategic part of any future solution to be adopted at a regional level. As a step in this direction, a packaging for the transport of irradiated fuel for MTR and TRIGA research reactors was designed by the tri-national team and a half-scale model equipped with the MTR version of the internal basket was constructed in Argentina and Brazil and tested in Brazil. Three test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. After failing the tests in the first two test series, the specimen successfully underwent the last test sequence. A second specimen, incorporating the structural improvements in view of the previous tests results, will be tested in the near future. Numerical simulations of the free drop and thermal tests are being carried out in parallel, in order to validate the computational modeling that is going to be used as a support for the package certification. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Nuclear Fuel Traces Definition in Storage Ponds of Research VVR-2 and OR Reactors in NRC 'Kurchatov Institute'

The Gas Plant complex is the experimental base of the Institute of Nuclear Reactors, which is part of the Kurchatov Institute. In 1954 the commissioning of the first Soviet water-cooled water-moderated research reactor VVR-2 on enriched uranium, and until 1983 the complex operated two research water-cooled water-moderated reactors 3 MW (VVR-2) and 300 kW (OR) capacity, which were dismantled in connection with the overall upgrades of the complex. The complex has three storage ponds in the reactor building. They are sub-surface vessels filled with water (the volume of water in each is about 6 m{sup 3}). In 2007-2013 the spent nuclear fuel from storages was removed for processing to 'Mayk'. Survey of Storage Ponds by Underwater Collimated Spectrometric System shows a considerable layer of slime on the bottom of ponds and traces of spent nuclear fuel in one of the storage. For determination qualitative and the quantitative composition of radionuclide we made complex α-, β-, γ- spectrometric research of water and bottom slimes from Gas Plant complex storage ponds. We found the spent nuclear fuel in water and bottom slime in all storage ponds. Specific activity of radionuclides in the bottom slime exceeded specific activity of radionuclides in the ponds water and was closed to levels of high radioactive waste. Analysis of the obtained data and data from earlier investigation of reactor MR storage ponds showed distinctions of specific activity of uranium and plutonium radionuclides. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Resourcing Strategies for the NRU Reactor Shutdown Planning and Execution - 19218

The National Research Universal Reactor (NRU) was commissioned in 1957. The NRU was fundamental for the development of the Candu reactors and, until 2016, quietly responsible for producing 40 percent of the world's supply of medical isotopes for diagnosis and cancer therapy. The Government of Canada decided in 2015 to discontinue operation as of March 31, 2018 due to challenges associated with running a 60-year old nuclear reactor. Anticipating this closing date allowed the Permanent Safe Shutdown Project (PSSP) to transition the NRU from an operating reactor to a state suitable for Storage with Surveillance. One of the biggest challenges to address was human resource management, which was rightly identified as the key to keeping the project within schedule and budget. The dilemma was, how to keep the work force motivated to contribute with the preparation of the shutdown while addressing job security concerns. This paper summarises Canadian Nuclear Laboratories (CNL) challenges and strategies to secure specialized subject matter experts, trades and operators for planning and executing the NRU reactor shutdown Project. The discussion also includes the solutions and lessons learned to date. (author)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Verification of Reactor Activation Modelling with Physical Characterization Data and the Impact on Long Term Assessments - 20303

Whiteshell Laboratories, located in Manitoba, Canada, provided research facilities for the Canadian nuclear industry since the early 1960's. The facility was centered on Whiteshell Reactor 1 (WR-1), an organically cooled, heavy water moderated nuclear reactor. WR-1 was safely shut down in 1985, defueled and drained, and has been in a safe Storage with Surveillance (Deferred Decommissioning) state until present day. In support of an Environmental Assessment for proposed in-situ disposal of WR-1, detailed characterization information regarding the reactor core components (calandria, calandria tubes, and pressure tubes) was necessary to provide confidence that the long-term impacts of the facility can be assessed adequately. A simplified estimate of the total activation of core components was completed in 1992 using conservative assumptions, expected reactor configuration and computer models built with WIMS-CRNL, ONEDANT and ORIGEN-S. Due to the simplification, and age of the work, there was considerable doubt that the modelled inventory estimate of the core components was sufficient to support long-term performance modeling of the planned in-situ decommissioning. Additional data would be required to provide this confidence. Several options were considered for characterizing the WR-1 core components, but it was decided to collect only a limited number of samples in an effort to validate the 1992 simplified estimate as bounding. The limited number of samples reduced risks to workers, both radiological and non-radiological, and significantly reduced the costs required to confidently define and bound the radiological inventory of activation products in the core. The results of the sampling were compared to the simplified activation estimation. There was good agreement between the modelled and sampled results, with all major contributors to total activity showing similar relative quantities. The samples were generally lower than predicted by the models, with few exceptions. Notable exceptions included Nb-94, which were higher in the sample results than the model. This was determined to be caused by the use of Niobium containing steel alloys in the stainless steel pressure tube that were not accounted for in the models. The comparison of results provided confidence that the original inventory estimates produced through the simplified model are conservative and bounding, and therefore suitable for use in the long-term assessment of in situ disposal of the WR-1 reactor. This paper provides details of the simplified computer model results, their comparison to results of the sampling, and the changes adopted in the assessment of the in-situ decommissioning of WR-1 as a result. (authors)

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

Recent research activities on high temperature pebble bed reactors at PSI - 18452

Recently, Paul Scherrer Institut (PSI) commenced a project on Feasibility and plausibility of innovative reactor concepts in an European electricity supply environment. The investigation in the project will be done on two reactor types: the pebble bed High-Temperature Gas-Cooled Reactor (HTGR) and Molten Salt Reactor (MSR) with the main focus being on the HTGR. Firstly, the project consists of an extensive literature study on the current status of the technology and research on the HTGR and MSR. Secondly, based on the findings from the literature survey, several topics of particular interest have been chosen for more thorough investigations. On the pebble bed HTGRs, the areas of investigation include risk assessment study, preliminary economic assessment, investigation of accident scenarios, and preliminary study on reduction of waste volumes. For both HTGR and MSR an analysis and optimization of the fuel cycle is foreseen. At the initial stage the equilibrium closed Th-U and U-Pu cycle performance will be evaluated and compared with 16 other reactor types. Later the knowledge from equilibrium study will be applied to optimize especially the MSR but also HTGR fuel cycle. At the end of the project, based on the literature survey and the results from the scoping studies, future research needs and areas of special interest will be identified for successive projects related to HTGR and MSR in PSI. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Experimental study on a simplified facility of HTGR reactor building response to depressurization accidental scenarios - 18516

The analysis and understanding of air ingress events is an important aspect in the design of The High Temperature Gas Cooled Reactor (HTGR) accident scenarios including Depressurized Loss of Forced Cooling (D-LOFC) events that allow for the possibility of air ingress into the Reactor Pressure Vessel (RPV) as the result of a break in the helium pressure boundary which can ultimately result in oxidation of the fuel elements and other nuclear-grade graphite components. In order to characterize the air ingress into the vented low pressure compartments (VLPC) of NGNP HTGR during hypothetical small break D-LOFC accidents experimentally, a 1/28 down scaled simplified reactor building model was developed. A non-dimensional similarity approach was employed for the scaling analysis of the experimental facility. Two experiments were conducted under similar experimental conditions, with 1.1 kg/h helium injection through the top of the reactor cavity, resulting in oxygen concentrations reaching to steady state within 4 minutes. Differences between the steady state oxygen concentrations results in the steam generator cavity of the two tests with the same experimental conditions were 0.51{+-}0.086% at the top of the steam generator cavity and 0.29{+-}0.11% at the bottom of the steam generator cavity. Difference in the pressure in the reactor between the two experiments were 0.09{+-}0.026 psig. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Design Report of the Molten-Chloride High Temperature Reactor (M-CHTR)

Liquid fueled molten salt reactors have seen renewed attention from reactor vendors and utility companies. In January of 2016, the Department of Energy awarded Southern Company, partnered with TerraPower, Electric Power Research Institute, Vanderbilt University, and Oak Ridge National Laboratory, a $40 million grant to support the development of the Molten Chloride Fast Reactor. This design report will discuss the design of the 2000 MW{sub th} Molten-Chloride High Temperature Reactor (M-CHTR). The M-CHTR is a liquid fuel molten sodium chloride fast reactor that is designed to consume spent nuclear fuel with normal operating temperatures near 1000 deg. C. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Transition to a U/Pu Fuel Cycle with Fast and Thermal Reactors

The purpose of this study was to analyze the transition from the current U.S. nuclear system to an advanced fuel cycle involving U/Pu continuous recycle in a fleet of fast and thermal reactors. This fuel cycle, known as 'EG29' in the 2014 Nuclear Fuel Cycle Evaluation and Screening (E and S), was identified as one of the four most promising out of a comprehensive list of fuel cycles that were grouped based on similar physics and performance characteristics. The EG29 fuel cycle involves continuously recycling the co-extracted U/Pu from a fleet of fast reactors represented by sodium-cooled fast reactors (SFRs) and burning the surplus amounts in a fleet of thermal reactors represented by pressurized water reactors with full mixed oxide (U,Pu)O{sub 2} cores (MOX-PWRs). The target energy generation ratio of SFRs to MOX-PWRs for the final EG29 system is 70:30. A 1% annual growth in nuclear energy demand was also assumed to be met during the entire scenario, including after EG29 is fully deployed. Therefore, to maintain system-wide mass balance, the SFR was designed to have a Pu inventory ratio (discharged/charged) of 1.4 compared to the MOX-PWR's 0.7. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Small CaH{sub 2} Moderated Thermal Reactor for Surface Power Generation I: Conceptual Design

National Aeronautics and Space Administration (NASA) published 'Journey to Mars' on October 8, 2015, stating their plan to send human to Mars for exploration, scientific research and colonization of Mars. Mission time on Martian surface is supposed to be ∼1 year, therefore, a surface power generation system is required for human habitation on Mars. Even though solar power and chemical battery have been used for the past missions, the former cannot generate power during month-scale dust storm on Mars, and the latter is not reliable for long-term use under a cold environment. On the other hand, nuclear power provides stable power even on Mars surface and, since kilowatt class power system is required in the planned mission, a fission power system is better than a radioisotope battery. An 8-year operation is required for the Mars surface power generation system with electrical power of 10 to 100 kWe and specific power of >5 We/kg, which means the system weight needs to be below 2000 kg in the case of 10 to 100 kWe system. In the current study, stretch goals for surface fission power system were set as follows: electric power of 20 kWe and total system weight below 1000 kg. In addition, nuclear security and critical safety are essential features for the surface fission power system considering launch accidents. A sufficient level of nuclear security is needed to prevent terrorists from obtaining nuclear materials even the reactor falls in a geographic area dominated by terrorists. From this view point, it is preferable not to use highly enriched uranium with a space reactor that can be diverted to nuclear weapons. Also, a nuclear reactor core should not reach criticality during a launch accident where the reactor core can be compressed by collision with sea or land. A novel concept of surface power reactor system that satisfies the above requirements is proposed in the current study. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Neutron Resonance Theory for Nuclear Reactor Applications: Modern Theory and Practices

The neutron resonance phenomena constitute one of the most fundamental subjects in nuclear physics as well as in reactor physics. It is the area where the concepts of nuclear interaction and the treatment of the neutronic balance in reactor fuel lattices become intertwined. The latter requires the detailed knowledge of resonance structures of many nuclides of practical interest to the development of nuclear energy. The most essential element in reactor physics is to provide an accurate account of the intricate balance between the neutrons produced by the fission process and neutrons lost due to the absorption process as well as those leaking out of the reactor system. The presence of resonance structures in many major nuclides obviously plays an important role in such processes. There has been a great deal of theoretical and practical interest in resonance reactions since Fermi’s discovery of resonance absorption of neutrons as they were slowed down in water. The resonance absorption became the center of attention when the question was raised as to the feasibility of the self-sustaining chain reaction in a natural uranium-fueled system. The threshold of the nuclear era was crossed almost eighty years ago when Fermi and Szilard observed that a substantial reduction in resonance absorption is possible if the uranium was made into the form of lumps instead of a homogeneous mixture with water. In the West, the first practical method for estimating the resonance escape probability in a reactor cell was pioneered by Wigner et al in early forties. (author)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗