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.

Evaluation of Radiation Effects on Residents Living Around the NSRR Under External Hazards

2020 ◽  
Vol 6 (2) ◽  
Author(s):  
Yuiko Motome ◽  
Yoshiya Akiyama ◽  
Hiroyuki Murao
Keyword(s):  
Radiation Effects ◽  
Research Reactor ◽  
Reactor Core ◽  
Fission Products ◽  
Internal Exposure ◽  
Spent Fuel ◽  
Loss Of Function ◽  
Safety Research ◽  
Spent Fuel Storage ◽  
Fuel Storage

Abstract The nuclear safety research reactor (NSRR) is a research reactor of training research isotopes general atomics—annular core pulse reactor (TRIGA-ACPR) type, located in the Nuclear Science Research Institute (NSRI). The NSRR facility has been utilized for fuel irradiation experiments to study the behaviors of nuclear fuels under reactivity-initiated accident (RIA) conditions. Under the new regulation standards, which was established after the Fukushima Daiichi accident, research reactors are regulated based on the risk of the facilities. The graded approach is introduced in the regulation. To apply the graded approach, the radiation effects on residents living around the NSRR under the external hazards were evaluated, and the level of the risk of the NSRR facility was investigated. This paper summarizes the result of the evaluation in the case where the safety functions are lost due to a tornado, an earthquake followed by a tsunami. There is fuel in the reactor core, fresh fuel storage, and spent fuel storage. As the effects from reactor core, we evaluate the external exposure to radiation and exposure from the release of fission products assuming that loss of function to shut down the reactor, break of cladding tubes, loss of reactor pool water, and collapse of the reactor building. As the effects from fresh fuel storage, we evaluate the internal exposure assuming that the fresh fuel particles released into the air because of breaking into pieces. In addition, we evaluate the critical safety assuming that the critical safety shapes of the fresh fuel storage are lost. As the effects from spent fuel storage, we evaluate the critical safety assuming that the critical safety shapes of the spent fuel storage are lost. All in all, the risk is confirmed to be relatively low, since the effective dose on the residents is found to be below 5 mSv per event due to the loss of the safety functions caused by the tornado, earthquake, and the accompanying tsunami.

scholarly journals Core parametric study for enhancing the radioisotope production in the IEA-R1 research reactor

2021 ◽  
Vol 9 (2B) ◽  
Author(s):  
Giovanni Laranjo Stefani ◽  
Frederico Antônio Genezini ◽  
Thiago Augusto Santos ◽  
João Manoel de Losada Moreira
Keyword(s):  
Neutron Flux ◽  
Parametric Study ◽  
Research Reactor ◽  
Reactor Core ◽  
Radioisotope Production ◽  
Spent Fuel ◽  
The Core ◽  
Safety Parameters ◽  
Fuel Elements ◽  
Average Neutron

In this work a parametric study was carried to increase the production of radioisotopes in the IEA-R1 research reactor. The changes proposed to implement in the IEA-R1 reactor core were the substitution of graphite reflectors by beryllium reflectors, the removal of 4 fuel elements to reduce the core size and make available 4 additional locations to be occupied by radioisotope irradiation devices. The key variable analyzed is the thermal neutron flux in the irradiation devices.  The proposed configuration with 20 fuel elements in an approximately cylindrical geometry provided higher average neutron flux (average increment of 12.9 %) allowing higher radioisotope production capability. In addition, it provided 4 more positions to install  irradiation devices which allow a larger number of simultaneous irradiations practically doubling the capacity of radioisotope production in the IEA-R1 reactor. The insertion of Be reflector elements in the core has to be studied carefully since it tends to promote strong neutron flux redistribution in the core. A verification of design and safety parameters of the proposed  core was carried out. The annual fuel consumption will increase about 17 % and more storage space for spent fuel will be required.   

Radiometric analysis of the spent fuel pool water and reactor coolant of ET-RR.1

2000 ◽  
Vol 88 (2) ◽  
Author(s):  
S.A.M. El-Mongy ◽  
M.S. El-Tahawy
Keyword(s):  
Fission Product ◽  
Research Reactor ◽  
Reactor Core ◽  
Spent Fuel ◽  
Reactor Coolant ◽  
Radiometric Analysis ◽  
Pool Water ◽  
Spent Fuel Pool ◽  
Fluorimetric Analysis ◽  
Radioactivity Levels

This work aims at analysis of radioactivity levels in the water of spent fuel pool and reactor core of the Egyptian 2MW research reactor (ET-RR.1 at Inshas). Gamma spectrometric and laser fluorimetric analysis have been used for carrying out this study. The fission product

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.

Analysis of Flow Blockage of a Single Fuel Assembly in the JRR-3 20MW Research Reactor

2018 ◽  
Author(s):  
Yuchuan Guo ◽  
Guanbo Wang ◽  
Dazhi Qian ◽  
Heng Yu ◽  
Bo Hu
Keyword(s):  
Fuel Assembly ◽  
System Analysis ◽  
Research Reactor ◽  
Natural Circulation ◽  
Reactor Core ◽  
Nucleate Boiling ◽  
Regulation System ◽  
Lower Power ◽  
Emergency Shutdown ◽  
Flow Blockage

The case of flow blockage of a single fuel assembly in the JRR-3 20MW open-pool-type research reactor is investigated without taking into account the effect of the power regulation system. The coolant system and multi-channel reactor core are modeled in detail using thermal hydraulic system analysis code RELAP5/MOD3.4. MDNBR (Minimum Departure From Nucleate Boiling Ratio) and the maximum fuel central temperature are investigated to assess the integrity of fuels. The fuel plates in blocked assembly are not damaged until the blockage ratio exceeds 70%. In addition, the mitigative effect of the assumed 18 MW lower power emergency shutdown operation on the accident is also discussed qualitatively. Results indicate that although the assumed lower power emergency shutdown operation cannot avoid the most severe operating condition, it can obviously mitigate the consequences of the accident. The reactor eventually remains in the long-term safe state when natural circulation is established.

A Research Reactor Source Term and Dose Results in the Event of a LOCA

2014 ◽  
Author(s):  
Charalampos Pappas ◽  
Andreas Ikonomopoulos ◽  
Athanasios Sfetsos ◽  
Spyros Andronopoulos ◽  
Melpomeni Varvayanni ◽  
...  
Keyword(s):  
Research Reactor ◽  
Source Term ◽  
Dispersion Model ◽  
Ventilation System ◽  
Atmospheric Dispersion ◽  
Reactor Core ◽  
Forecast Model ◽  
Scale Model ◽  
Continuous Reactor ◽  
Reactor Building

The present study discusses the source term derivation and dose result calculation for a hypothetical accident sequence in the Greek Research Reactor – 1 (GRR-1). A loss-of-coolant accident (LOCA) has been selected as a credible accident sequence. The source term derivation has been based on the GRR-1 confinement performance where the inventory has been computed assuming continuous reactor operation. A core damage fraction of 30% has been considered for the calculations while conservative core release fractions have been employed. The radionuclides released from the reactor core to the confinement atmosphere have been subjected to natural decay, deposition on and resuspension from various internal surfaces before being led to the release pathway. It has been assumed that an emergency shutdown is initiated immediately after the beginning of the accident sequence and the emergency ventilation system is also activated. Subsequently, the source term has been derived comprising of noble gases, iodine and aerosol. The JRODOS computational software for off-site nuclear emergency management has been utilized to estimate the dose results from the LOCA-initiated source term that is released in its entirety from the reactor stack at ambient temperature. The Local Scale Model Chain in conjunction with the DIPCOT atmospheric dispersion model that is embedded in JRODOS have been used with proper parameterization of the calculation settings. Five weather scenarios have been selected as representative of typical meteorological conditions at the reactor site. The scenarios have been assessed with the use of the Weather Research and Forecast model. Total effective, skin, thyroid, lung and inhalation doses downwind of the reactor building and up to a distance of 10 km have been calculated for each weather scenario and are presented. The total effective gamma dose rate at a fixed distance from the reactor building has been assessed. The radiological consequences of the dose results are discussed.

scholarly journals Study of corrosion of aluminum alloys of nuclear purity in ordinary water: Part two

2005 ◽  
Vol 20 (2) ◽  
pp. 45-60 ◽  
Author(s):  
Milan Pesic ◽  
Tatjana Maksin ◽  
Gabrijela Jordanov ◽  
Rajko Dobrijevic ◽  
Zoja Idjakovic
Keyword(s):  
Aluminum Alloys ◽  
International Atomic Energy Agency ◽  
Research Reactor ◽  
Atomic Energy ◽  
Spent Fuel ◽  
Research Project ◽  
Energy Agency ◽  
Ordinary Water ◽  
Fuel Storage ◽  
Frame Work

Since 2002, the effects of corrosion on aluminum alloys of nuclear purity in ordinary water of the spent fuel storage pool of the RA re search reactor at VINCA Institute of Nuclear Sciences have been examined in the frame work of the International Atomic Energy Agency Coordinated Research Project "Corrosion of Research Reactor Aluminum Clad Spent Fuel in Water". Coupons were ex posed to the pool water for a period of six months to six years. The second part of this study comprises extensive results obtained by detailed visual and microscopic examinations of the surfaces of the coupons and represents an integral part of the first report on the topic, previously presented in this journal.

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