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

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

Neutronics Optimization of UN Fuel Pin Assemblies for a Sodium-Cooled, Small Modular Reactor

2011 ◽  
Author(s):  
Timothy M. Schriener ◽  
Mohamed S. El-Genk
Keyword(s):  
Thermal Power ◽  
Reactor Core ◽  
Central Cavity ◽  
Preliminary Results ◽  
The Core ◽  
Small Modular Reactor ◽  
Fuel Assemblies ◽  
Fuel Pin ◽  
Reference Design ◽  
Centerline Temperature

This paper presents preliminary results of neutronics and thermal-hydraulics design analysis of a sodium cooled, small modular reactor (SMR). The reactor’s nominal thermal power is 150 MWth at sodium inlet and exit temperatures of 630 and 780 K. The reactor core is comprised of three rings of shrouded hexagonal assemblies of 19.8% enriched UN fuel pins and a hexagonal assembly of enriched B4C pins in the central cavity for a coarse reactivity control. The objectives are to provide enough excess reactivity for achieving a refueling cycle > 5 year, maintaining a more even coolant flow in the core assemblies and keeping the peak centerline temperature of UN fuel pins < 1300 K. Fuel assemblies with scalloped shroud walls, 4 rings and 1.942 cm diameter fuel pins with p/d = 1.098 are selected for the reference design of the present SMR. In this design, peak fuel centerline temperature is only 1240 K and the beginning-of-life, cold-clean excess reactivity is $26.67.

Application of LAPUR6 to Lungmen ABWR Stability Analysis

2013 ◽  
Vol 284-287 ◽  
pp. 1146-1150 ◽  
Author(s):  
Hao Tzu Lin ◽  
Jong Rong Wang ◽  
Chun Kuan Shih
Keyword(s):  
Stability Analysis ◽  
Nuclear Power ◽  
Thermal Power ◽  
Core Stability ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Design Features ◽  
Water Reactor ◽  
The Core ◽  
Stability Performance

Lungmen nuclear power plant (NPP) is the first ABWR (Advanced Boiling Water Reactor) in Taiwan and still under construction. It has two identical units with 3,926 MWt rated thermal power each and 52.2×106 kg/hr rated core flow. The core has 872 bundles of GE14 fuel, and the steam flow is 7.637×106 kg/hr at rated power. According to the chapter 4 of Lungmen NPP FSAR (Final Safety Analysis Report), the design features of Lungmen NPP improve the core stability performance and assure that it is more stable than the current BWR (Boiling Water Reactor) NPP in the normal operating regions. In this research, the LAPUR6 stability analysis of Lungmen NPP is performed in order to versify the design features of Lungmen NPP which causes the more stable than the current BWR NPPs. The analysis results of LAPUR6 indicate that the design features of Lungmen NPP can improve the core stability performance effectively and result in the more stable than the current BWR NPPs.

scholarly journals Possibility of Simulating Forced Convection in Fast Neutron Reactors Using a Light Water Test Facility

2018 ◽  
Vol 3 (3) ◽  
pp. 279
Author(s):  
V.I. Slobodchuk ◽  
E.A. Avramova ◽  
E.M. Shchennikova ◽  
D.A. Shal'kov
Keyword(s):  
Heat Transfer ◽  
Nuclear Power ◽  
Similarity Theory ◽  
Test Facility ◽  
Test Model ◽  
Nuclear Power Reactor ◽  
Light Water ◽  
Reactor Installation ◽  
Water Test ◽  
Metal Coolant

The paper evaluates the possibility of modeling the heat transfer phenomena in a liquid-metal coolant using a light water test facility. A large nuclear power reactor (like the BN-1200 project) was selected as a reactor installation to be modeled. To validate the model, the similarity theory and the “black box” method were used. The paper uses the experience of a number of researchers in this field, in particular, the accepted assumptions which do not result in serious loss in modeling accuracy. The governing criteria of similarity were estimated based on the fundamental differential equations of convective heat transfer, so were the conditions under which it is possible to model sodium coolant by using  light water with adequate accuracy. The paper presents the scales of the parameters used for the model - reactor comparison. Dependence curves of certain scales with regard to others are constructed, and the possibility of achieving similarity of certain parameters in modeling was estimated. Recommendations are provided on designing a water test model of the BN reactor and on carrying out experiments using this test model.

scholarly journals Perhitungan Perubahan Reaktivitas Pada Reaktor Serba Guna G.A. Siwabessy Akibat Pengoperasian Power Ramp Test Facility

2019 ◽  
Vol 4 ◽  
pp. 121
Author(s):  
A Suparmi ◽  
Tuti Dwi Setyaningsih ◽  
Suharyana Suharyana ◽  
Fuad Anwar ◽  
Riyatun Riyatun
Keyword(s):  
Reactor Core ◽  
Test Facility ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Ramp Test ◽  
Fuel Pin ◽  
Safe Limit ◽  
Reactor Operating ◽  
Test Fuel

<p><strong>Abstract: </strong>Power Ramp Test Facility (PRTF) is one of the irradiation facility contained in the Multipurpose Reactor GA Siwabessy. This facility is used to test the reactor fuel element pin-type Pressurized Water Reactor. As a result of the entry of foreign bodies cause changes reactor conditions, one of which is expressed with the amount of reactivity to assess the safety of the reactor due to the operation PRTF. PRTF operation simulation and calculation is done using software neutronics MCNP6. Test UO2 fuel enriched assumed at 5% with constant power reactor operating at 15 MW and test fuel pin placed on PRTF within 0, 20, 40, 60, 80, 100, 120, and 140 mm from the centre of the reactor core. Change of reactivity values required in order to secure the reactor, maximal value is 0,5%<em></em>.  The calculation were obtained at each position is (<em></em><em></em>;  <em></em>;  <em></em>; <em></em>;<em></em>; <em></em>; <em></em>; <em></em>). Change of reactivity values smaller than the safe limit. Therefore, the study of reactivity changes PRTF operation to test fuel pin is secure.</p><p><strong>Abstrak: </strong>Power Ramp Test Facility (PRTF) merupakan salah satu fasilitas iradiasi yang terdapat pada Reaktor Serba Guna G.A. Siwabessy. Fasilitas ini digunakan untuk menguji pin elemen bahan bakar reaktor tipe Pressurized Water Reactor. Akibat dari masuknya benda asing menyebabkan perubahan kondisi reaktor, salah satunya dinyatakan dengan besaran reaktivitas untuk mengkaji keselamatan reaktor akibat pengoperasian PRTF. Simulasi pengoperasian PRTF dan perhitungan netronik dilakukan menggunakan perangkat lunak MCNP6. Bahan bakar uji UO2 diasumsikan diperkaya sebesar 5% dengan daya operasi reaktor konstan sebesar 15 MW. Pin bahan bakar uji diletakkan pada PRTF berjarak 0, 20, 40, 60, 80, 100, 120, dan 140 mm dari arah pusat teras reaktor. Nilai perubahan reaktivitas yang dipersyaratkan agar reaktor aman adalah , sedangkan nilai perubahan reaktivitas dari penelitian pada masing-masing posisi dari pusat reactor adalah (;  ;  ; ;; ; ; ) . Nilai perubahan reaktivitas akibat masuknya pin bahan bakar di PRTF mempunyai nilai perubahan reaktivitas 1/10 kali lebih kecil daripada batas aman. Oleh karena itu, ditinjau dari kajian  nilai perubahan reaktivitas maka pengoperasian PRTF untuk uji pin bahan bakar adalah aman.</p>

scholarly journals Concept of a Fast Breeder Reactor To Transmute MA and LLFP

2021 ◽  
Author(s):  
Toshio Wakabayashi
Keyword(s):  
Radioactive Waste ◽  
Power Systems ◽  
Fast Reactor ◽  
Nuclear Power ◽  
Fission Products ◽  
Fast Breeder Reactor ◽  
Nuclear Power Reactor ◽  
The Core ◽  
Effective Use

Abstract The long-term issues of nuclear power systems are the effective use of uranium resources and the reduction of radioactive waste. Important radioactive wastes are minor actinides (MA: 237 Np, 241 Am, 243 Am, etc.) and long-lived fission products (LLFP: 129 I, 99 Tc, 79 Se, etc.). The purpose of this study was to show a system that can simultaneously achieve the breeding of fissile materials and the transmutation of MA and LLFP in one fast reactor. Transmutation was carried out by loading innovative Duplex type MA fuel in the core region and LLFP containing moderator in the first layer of the radial blanket. Breeding was achieved in the core and axial blanket. As a result, it was clarified that in this fast breeder reactor, a breeding ratio of about 1.1 was obtained, and MA and LLFP achieved a support ratio of 1 or more. The transmutation rate was 10.3%/y for 237 Np, 14.1%/y for 241 Am, 9.9%/y for 243 Am, 1.6%/y for 129 I, 0.75%/y for 99 Tc, and 4%/y for 79 Se. By simultaneously breeding fissile materials and transmuting MA and LLFP in one fast reactor, it will be possible to solve the long-term issues of the nuclear power reactor system, such as securing nuclear fuel resources and reducing radioactive waste.

Station Blackout Analysis of Lungmen ABWR Using TRACE

2012 ◽  
Author(s):  
Jong-Rong Wang ◽  
Hao-Tzu Lin ◽  
Hsiung-Chih Chen ◽  
Wei-Chen Wang ◽  
Chunkuan Shih
Keyword(s):  
Nuclear Power ◽  
Thermal Power ◽  
Electrical Power ◽  
Heat Removal ◽  
Fuel Temperature ◽  
Core Flow ◽  
The Core ◽  
Ac Power ◽  
Station Blackout ◽  
Core Cooling

The Lungmen NPP is the first ABWR (Advanced Boiling Water Reactor) nuclear power plant in Taiwan, consisting of two identical units with 3,926 MWt rated thermal power each and 52.2×106 kg/h rated core flow. The core of Lungmen NPP has 872 bundles of GE14 fuel. There are 10 reactor internal pumps (RIP) in the reactor vessel, providing 111% rated core flow at the nominal operating speed of 151.84 rad/sec. A station blackout (SBO) is defined as the loss of offsite electrical power concurrent with turbine trip and unavailability of the onsite emergency AC power. These result in the loss of core cooling and heat removal systems that rely on the above AC power for their operation. In this research, the TRACE SBO model of Lungmen ABWR has been developed in order for the analysis of SBO transient. The initial condition of SBO transient is 100% rated power/100% rated core flow. The TRACE’s results show that the reactor fuel temperature has been reached 1088.71 K (the zirconium-water reaction may generate) at about 3200 sec. It indicates that the fuels might be damaged after 3200 sec if the RCIC and ACIWA failed to activate in this transient.

Pebble Bed Modular Reactor Fuel Enrichment Discrimination Using Delayed Neutrons

2008 ◽  
Author(s):  
R. Skoda ◽  
J. Rataj ◽  
J. Uher
Keyword(s):  
Nuclear Power ◽  
Power Reactor ◽  
Theoretical Calculations ◽  
Nuclear Power Reactor ◽  
Pebble Bed ◽  
Delayed Neutrons ◽  
The Core ◽  
Start Up ◽  
On Line ◽  
First Time

The Pebble Bed Modular Reactor (PBMR) is a helium-cooled, graphite-moderated high temperature nuclear power reactor which utilise fuel in form of spheres that are randomly loaded and continuously circulated through the core until they reach their prescribed end-of-life burn-up limit. When the reactor is started up for the first time, the lower-enriched start-up fuel is used, mixed with graphite spheres, to bring the core to criticality. As the core criticality is established and the start-up fuel is burned-in, the graphite spheres are progressively removed and replaced with more start-up fuel. Once it becomes necessary for maintaining power output, the higher enriched equilibrium fuel is introduced to the reactor and the start-up fuel is removed. During the initial run of the reactor it is important to discriminate between the irradiated startup fuel and the irradiated equilibrium fuel to ensure that only the equilibrium fuel is returned to the reactor. There is therefore a need for an on-line enrichment discrimination device that can discriminate between irradiated start-up fuel spheres and irradiated equilibrium fuel spheres. The device must also not be confused by the presence of any remaining graphite spheres. Due to it’s on-line nature the device must accomplish the discrimination within tight time limits. Theoretical calculations and experiments show that Fuel Enrichment Discrimination based on delayed neutrons detection is possible. The paper presents calculations and experiments showing viability of the method.

scholarly journals Concept of a fast breeder reactor to transmute MAs and LLFPs

2021 ◽  
Vol 11 (1) ◽  
Author(s):  
Toshio Wakabayashi
Keyword(s):  
Radioactive Waste ◽  
Power Systems ◽  
Fast Reactor ◽  
Nuclear Power ◽  
Fission Products ◽  
Fast Breeder Reactor ◽  
Nuclear Power Reactor ◽  
The Core ◽  
Effective Use

AbstractThe long-term issues of nuclear power systems are the effective use of uranium resources and the reduction of radioactive waste. Important radioactive wastes are minor actinides (MAs: 237Np, 241Am, 243Am, etc.) and long-lived fission products (LLFPs: 129I, 99Tc, 79Se, etc.). The purpose of this study was to show a concept that can simultaneously achieve the breeding of fissile materials and the transmutation of MAs and LLFPs in one fast reactor. Transmutation was carried out by loading innovative Duplex-type MA fuel in the core region and LLFP-containing moderator in the first layer of the radial blanket. Breeding was achieved in the core and axial blanket. As a result, it was clarified that in this fast breeder reactor, a breeding ratio of approximately 1.1 was obtained, and MAs and LLFPs achieved a support ratio of 1 or more. The transmutation rate was 10.3%/y for 237Np, 14.1%/y for 241Am, 9.9%/y for 243Am, 1.6%/y for 129I, 0.75%/y for 99Tc, and 4%/y for 79Se. By simultaneously breeding fissile materials and transmuting MAs and LLFPs in one fast reactor, it will be possible to solve the long-term issues of the nuclear power reactor system, such as securing nuclear fuel resources and reducing radioactive waste.

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