Reactor Shutdown and Reactor Restart

Fundamentals of CANDU Reactor Physics ◽

10.1115/1.884836_ch10 ◽

2021 ◽

pp. 101-112

Author(s):

Wei Shen ◽

Benjamin Rouben

Keyword(s):

Neutron Flux ◽

Thermal Power ◽

Power Measurement ◽

Reactor Physics ◽

Fission Rate ◽

Neutron Power ◽

Reactor Shutdown ◽

Ion Chambers

The power referred to most frequently in reactor physics is neutron power. Neutron power is essentially the fission rate multiplied by the average prompt energy released and recovered per fission (see Section 2.1.2). It is also called “prompt” power, as it appears very quickly following fission. We cannot measure neutron power directly, but we do monitor the neutron flux with ion chambers located outside the calandria and in-core flux detectors. These neutronic signals are calibrated to the thermal-power measurement which allows neutron power to be derived.

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The REBUS Experimental Programme for Burn-Up Credit

12th International Conference on Nuclear Engineering, Volume 3 ◽

10.1115/icone12-49164 ◽

2004 ◽

Author(s):

Pierre D’hondt ◽

Peter Baeten ◽

Leo Sannen ◽

Daniel Marloye ◽

Benoit Lance ◽

...

Keyword(s):

Water Level ◽

Nuclear Research ◽

Research Centre ◽

Reactor Physics ◽

Fission Rate ◽

Mox Fuel ◽

Experimental Programme ◽

Fuel Bundle ◽

International Programme ◽

Good Agreement

An international programme called REBUS for the investigation of the burn-up credit has been initiated by the Belgian Nuclear Research Centre SCK-CEN and Belgonucle´aire with the support of EdF and IRSN from France and VGB, representing German nuclear utilities and NUPEC, representing the Japanese industry. Recently also ORNL from the U.S. joined the programme. The programme aims to establish a neutronic benchmark for reactor physics codes in order to qualify the codes for calculations of the burn-up credit. The benchmark exercise investigates the following fuel types with associated burn-up: reference fresh 3.3% enriched UO2 fuel, fresh commercial PWR UO2 fuel and irradiated commercial PWR UO2 fuel (54 GWd/tM), fresh PWR MOX fuel and irradiated PWR MOX fuel (20 GWd/tM). The experiments on the three configurations with fresh fuel have been completed. The experiments show a good agreement between calculation and experiments for the different measured parameters: critical water level, reactivity effect of the water level and fission-rate and flux distributions. In 2003 the irradiated BR3 MOX fuel bundle was loaded into the VENUS reactor and the associated experimental programme was carried out. The reactivity measurements in this configuration with irradiated fuel show a good agreement between experimental and preliminary calculated values.

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Measurement of Neutron Flux Spectra by a Multiple Foil Activation Iterative Method and Comparison with Reactor Physics Calculations and Spectrometer Measurements

Nuclear Science and Engineering ◽

10.13182/nse69-a18853 ◽

1969 ◽

Vol 36 (1) ◽

pp. 15-27 ◽

Cited By ~ 20

Author(s):

W. N. McElroy ◽

S. Berg ◽

T. B. Crockett ◽

R. J. Tuttle

Keyword(s):

Iterative Method ◽

Neutron Flux ◽

Reactor Physics ◽

Foil Activation

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IRRADIATION CHARACTERISTIC OF NATURAL UO2 PIN PHWR TARGET AT PRTF OF RSG – GAS CORE

JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA ◽

10.17146/tdm.2017.19.2.3306 ◽

2017 ◽

Vol 19 (2) ◽

pp. 71

Author(s):

Jati Susilo ◽

Tagor Malem Sembiring ◽

Winter Dewayatna

Keyword(s):

Neutron Flux ◽

Nuclear Power ◽

Irradiation Time ◽

Thermal Power ◽

Analytical Tool ◽

Test Facility ◽

Nuclear Power Reactor ◽

Ramp Test ◽

The Core ◽

Fuel Pin

The RSG-GAS reactor has a facility for irradiation of the fuel pin of nuclear power reactor, namely Power Ramp Test Facility (PRTF). The in-house fabrication PWR fuel pin has prepared for irradiations in the PRTF facility, currently, while the various enrichments of uranium are analyzed using the analytical tool. In the next step, it is planned to perform an irradiation of PHWR fuel pin sample of natural UO2 in the facility. Before irradiation in the core, it should be analyzed by using the analytical tool. The objectives of this paper are to optimize irradiation time based on the burn-up, the generated linear power and the neutron flux level at the target. The 3-dimension calculations have been carried out by using the CITATION code in the SRAC2006 code system. Since the coolant of the reactor is H2O, the effect of moderators in the pressurized tube, H2O and D2O, were analyzed, as well as pellet radius and moderator densities. The calculation results show that the higher linear power as irradiation time longer is occurred preferably in the D2O moderator than in H2O. For the D2O moderator, the higher pressure affects the lower density and longer irradiation time. The maximum irradiation time for natural UO2 fuel pin with the pressurized D2O moderator is about 9.5×104 h, with the linear power of 700 W/cm. During irradiation, neutronic parameters of the core such as excess reactivity and ppf show a very small change, still far below design value.Keywords: PHWR, Neutron Flux, Thermal Power, PRTF, RSG-GAS KARAKTERISTIK IRADIASI TARGET PIN PHWR UO2 ALAM PADA PRTF TERAS RSG – GAS. Teras RSG-GAS dilengkapi dengan fasilitas untuk uji iradiasi bahan bakar nuklir atau disebut dengan Power Ramp Test Fasility (PRTF). Saat ini sedang dilpersiapkan untuk dilakukan uji sample pin bahan bakar PWR pada fasilitas PRTF. Analisis terhadap uji iradiasi sample pellet UO2 dengan berbagai pengkayaan telah dilakukan menggunakan paket program komputer. Dimasa yang akan datang, uji iradiasi pin bahan bakar PHWR UO2 alam juga sedang dalam perencanaan. Sebelum diiradiasi di dalam teras, maka terlebih dahulu harus dilakukan analisis dengan menggunakan paket program komputer. Tujuan dari penelitian ini adalah optimasi uji iradiasi pin bahan bakar UO2 alam sebagai fungsi waktu iradiasi berdasarkan burn-up, daya linier dan fluks neutron. Perhitungan teras RSG-GAS dilakukan dengan paket program SRAC2006 modul CITATION dalam bentuk geometri 3 dimensi. Analisis dilakukan terhadap pengaruh penggunaan jenis moderator pada tabung tekan iradiasi (H2O dan D2O), perubahan ukuran pelllet UO2 dan perubahan besarnya densitas moderator D2O. Dari analisis hasil perhitungan diketahui bahwa semakin lama waktu iradiasi akan menghasilkan daya termal yang semakin besar jika menggunakan moderator D2O dibandingkan H2O. Semakin tinggi tekanan atau semakin kecil densitas moderator, maka akan menghasilkan daya termal yang semakin besar seiring bertambah lamanya waktu iradiasi. Batas maksimal waktu iradiasi untuk pin bahan bakar UO2 alam dengan moderator D2O bertekanan adalah sekitar 9,5×104 jam, dengan batasan daya linier desain kemampuan peralatan, 700 W/cm. Selama iradiasi, nilai parameter neutronik teras reaktor seperti reaktivitas lebih dan ppf hanya menunjukkan perubahan yang sangat kecil, masih jauh dibawah batas yang ditetapkan dalam desain.Kata kunci: PHWR, Fluks Neutron, Daya Termal, PRTF, RSG-GAS

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Optimization of Nanosecond Fission Ion Chambers for Reactor Physics Applications

IEEE Transactions on Nuclear Science ◽

10.1109/tns.1975.4327729 ◽

1975 ◽

Vol 22 (1) ◽

pp. 686-690 ◽

Cited By ~ 6

Author(s):

N. W. Hill ◽

J. T. Mihalczo ◽

J. W. Allent ◽

M. M. Chiles

Keyword(s):

Reactor Physics ◽

Ion Chambers

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Neutron Activation Analysis of Some Building Material

JOURNAL OF ADVANCES IN PHYSICS ◽

10.24297/jap.v14i2.7507 ◽

2018 ◽

Vol 14 (2) ◽

pp. 5564-5573

Author(s):

Tarek Mohamed Talaat Salama ◽

N. A. Mansour ◽

M. Fayez-Hassan

Keyword(s):

Neutron Activation Analysis ◽

Activation Analysis ◽

Neutron Activation ◽

Neutron Flux ◽

Building Material ◽

Thermal Power ◽

Essential Elements ◽

Ratio Method ◽

Metrology In Chemistry ◽

Amount Of Substance

Neutron activation analysis (NAA), based on the comparator method, has the potential to fulfill the requirements of a primary ratio method as defined in 1998 by the Comité Consultatif pour la Quantité de Matière — Métrologie en Chimie (CCQM, Consultative Committee on Amount of Substance — Metrology in Chemistry Studies of Neutron Activation Analysis (NAA) have been carried out on different Egyptian building material samples. The technique of neutron activation analysis is based on the measurement of radiation released by the decay of radioactive nuclei formed by neutron irradiation of the material. The most suitable source of neutrons for such an application is usually a research reactor. The samples that can be analyzed with this method stem from a number of different fields, including medicine, nutrition, biology, chemistry, forensics, the environment and mining. Neutron activation analysis can be performed in a variety of ways. This depends on the element and the corresponding radiation levels to be measured, as well as on the nature and the extent of interference from other elements present in the sample. Most of the methods used are non-destructive, based on the detection of gamma radiation emitted by the irradiated material after or during the irradiation. Next to education and training, neutron activation analysis is the most widely used application of research reactors. Almost any reactor operating at 10-30 kilowatt of thermal power is capable of providing a sufficient neutron flux to irradiate samples for selective applications of this analysis technique. Another method of NAA by using two Am-Be isotopic neutron sources of activity 5 Ci were used in this investigation. The accomplished gamma rays were measured using 70 % HPGe spectrometer. This work demand to estimate the elements contained in cement products and its quality control. X-ray Fluorescence (XRF) measurements were done for confirming our results, and for determining the average neutron flux of 3.7× 103 n/cm2sec. The Natural radioactivities of these samples were measured before the analysis to know the background level of 40K, 238U and232Th nuclei. The results investigated that NAA agree with the results of XRF and the world range of the cement concentration of the essential elements Ca, Al, Na, Fe, Mn, V, Sr and Si.

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An Optimization Design of Collimator in The Thermal Column of Kartini Reactor For BNCT

Indonesian Journal of Physics and Nuclear Applications ◽

10.24246/ijpna.v2i2.54-64 ◽

2017 ◽

Vol 2 (2) ◽

pp. 54

Author(s):

M. Ibnu Khaldun ◽

Andang Widi Harto ◽

Yohannes Sardjono

Keyword(s):

Neutron Flux ◽

Neutron Beam ◽

Optimization Design ◽

Epithermal Neutron ◽

Thermal Power ◽

Epithermal Neutron Flux ◽

Aperture Diameter ◽

In Vivo Experiment ◽

Γ Ray

Studies were carried out to design a collimator which results in epithermal neutron beam for in vivo experiment of Boron Neutron Capture Therapy (BNCT) at the Kartini Research Reactor by means of Monte Carlo N-Particle (MCNP) codes. Reactor within 100 kW of thermal power was used as the neutron source. All materials used were varied in size, according to the value of mean free path for each material. MCNP simulations indicated that by using 6 cm thick of Natural Nickel as collimator wall, 65 cm thick of Al as moderator, 3 cm thick of Ni-60 as filter, 6 cm thick of Bi as γ-ray shielding, 3.5 cm thick of Li2CO3-polyethilene, with 2 cm aperture diameter. Epithermal neutron beam with maximum flux of 6.60 x 108n.cm-2.s-1 could be produced. The beam has minimum fast neutron and γ-ray components of, respectively, 1.82 x 10-13Gy.cm2.n-1 and 1.70 x 10-13 Gy.cm2.n-1, minimum thermal neutron per epithermal neutron ratio of 0.041, and maximum directionality of 2,12. It did not fully pass the IAEA’s criteria, since the epithermal neutron flux was below the recommended value, 1.0 x 109 n.cm-2.s-1. Nonetheless, it was still usable with epithermal neutron flux exceeding 5.0 x 108 n.cm-2.s-1. it is still feasible for BNCT in vivo experiment.

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Power Measurement Methodologies for Pool Nuclear Research Reactors

Latin American Journal of Development ◽

10.46814/lajdv3n2-032 ◽

2021 ◽

Vol 3 (2) ◽

pp. 882-892

Author(s):

Amir Zacarias Mesquita ◽

Daniel Artur Pinheiro Palma ◽

Hugo Cesar Rezende ◽

Alexandre Melo De Oliveira ◽

...

Keyword(s):

Energy Balance ◽

Fuel Element ◽

Nuclear Reactors ◽

Thermal Power ◽

Nuclear Research ◽

Reactor Power ◽

Power Measurement ◽

Fuel Temperature ◽

Power Monitoring ◽

On Line

Redundancy and diversity are two important criteria for power measurement in nuclear reactors. Other criteria such as accuracy, reliability and response speed are also of major concern. Power monitoring of nuclear reactors is normally done by means of neutronic instruments, i.e. by the measurement of neutron flux. The greater the number of channels for power measuring the greater is the reliability and safety of reactor operations. The aim of this research is to develop new methodologies for on-line monitoring of nuclear reactor power using other reliable processes. One method uses the temperature difference between an instrumented fuel element and the pool water below the reactor core. Another method consists of the steady-state energy balance of the primary and secondary reactor cooling loops. A further method is the calorimetric procedure whereby a constant reactor power is monitored as a function of the temperature-rise rate and the system heat capacity. Another methodology, which does not employ thermal methods, is based on measurement of Cherenkov radiation produced within and around the core. The first three procedures, fuel temperature, energy balance and calorimetric, were implemented in the IPR-R1 Triga nuclear research reactor at Belo Horizonte (Brazil) and are the focus of the work described here. Knowledge of the reactor thermal power is very important for precise neutron flux and fuel element burnup calculations. The burnup is linearly dependent on the reactor thermal power and its accuracy is important in the determination of the mass of burned 235U, fission products, fuel element activity, decay heat power generation and radiotoxicity. The thermal balance method developed in this project is now the standard methodology used for IPR-R1 Triga reactor power calibration and the fuel temperature measuring is the most reliable way of on-line monitoring of the reactor power. This research project primarily aims at increasing the reliability and safety of nuclear reactors using alternative methods for power monitoring.

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The REBUS Experimental Programme for Burn-Up Credit

10th International Conference on Nuclear Engineering, Volume 4 ◽

10.1115/icone10-22022 ◽

2002 ◽

Author(s):

Pierre D’hondt ◽

Klaas van der Meer ◽

Peter Baeten ◽

Daniel Marloye ◽

Benoit Lance ◽

...

Keyword(s):

Gamma Spectrometry ◽

Fission Products ◽

Nuclear Research ◽

Reactor Physics ◽

Direct Evaluation ◽

Fission Rate ◽

Mox Fuel ◽

Experimental Implementation ◽

International Programme ◽

Length Determination

An international programme called REBUS (REactivity tests for a direct evaluation of the Burn-Up credit on Selected irradiated LWR fuel bundles) for the investigation of the burn-up credit has been initiated by the Belgian Nuclear Research Center SCK•CEN and Belgonucle´aire with the support of USNRC, EdF from France, VGB, representing German nuclear utilities and NUPEC, representing the Japanese industry. The programme aims to establish a neutronic benchmark for reactor physics codes. This benchmark will qualify the codes to perform calculations of the burn-up credit. The benchmark exercise investigates the following fuel types with associated burn-up: • Reference 3.3% enriched UO2 fuel; • Fresh commercial PWR UO2 fuel; • Irradiated commercial PWR UO2 fuel (51 GWd/tM); • Fresh PWR MOX fuel; • Irradiated PWR MOX fuel (20 GWd/tM). Reactivity effects are measured in the critical facility VENUS. Fission rate and flux distributions in the experimental bundles will be determined. The accumulated burn-up of all rods is measured non-destructively in a relative way by gross gamma-scanning, while some rods are examined by gamma-spectrometry for an absolute determination of the burn-up. Some rods will be analyzed destructively with respect to accumulated burn-up, actinides content and TOP-19 fission products (i.e. those non-gaseous fission products that have most implications on the reactivity). Additionally some irradiated rods have undergone a profilometry and length determination. The experimental implementation of the programme has started in 2000 with major changes in the VENUS critical facility. Gamma scans, profilometry, length determination and gamma-spectrometry measurements on the MOX fuel have been performed. In the course of October 2001 the first fresh fuel configuration will be investigated. In the same period the commercial irradiated fuel will arrive at the SCK•CEN hot cells and will be refabricated into fuel rodlets of 1 meter length.

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