Use of erbium as a burnable absorber for the VVER reactor core life extension

The paper presents the results of a computational and theoretical analysis concerned with the use of erbium as a burnable absorber in VVER-type reactors. Partial refueling options for the reactor life extension to 18 and 24 months is considered, the refueling ratio being equal to three for the 18-month life and to two for the 24-month life. Erbium is expected to be present in all fuel elements in the FA with the same weight content. The influence of the erbium weight content on such neutronic characteristics of the reactor and fuel as burn-up, reactivity coefficients, residual volume of “liquid" control, and amounts of the liquid radioactive waste (LRW) formed was assessed. The calculations were performed using a simplified model of refueling without FA reshuffling. An infinite array of polycells consisting of FAs with different in-core times was considered. The escape of neutrons from the core was taken into account by selecting the critical value K∞ at the end of life. Erbium does not burn up in full for the lifetime which affects the fuel burn-up as compared with the liquid excessive reactivity compensation system. The reduction is 0.7% per 0.1% of the erbium weight load in the fuel elements. This, however, also reduces the maximum content of the boron absorber in the coolant and the LRW accumulation in the ratio of 5% per 0.1% of the erbium weight load. Erbium influences the spectral component of the coolant temperature reactivity coefficient which turns out to be negative even with its minor weight fraction in fuel elements, and a reduction in the boron absorber fraction leads to a positive value of the density reactivity coefficient. As a result, the overall coolant temperature reactivity coefficient has a negative value throughout the lifetime.

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Improved FAST Device for Inherent Safety of Oxide-Fueled Sodium-Cooled Fast Reactors

Energies ◽

10.3390/en14154610 ◽

2021 ◽

Vol 14 (15) ◽

pp. 4610

Author(s):

Ahmed Amin E. Abdelhameed ◽

Chihyung Kim ◽

Yonghee Kim

Keyword(s):

Operating Temperature ◽

Reactor Core ◽

Coolant Temperature ◽

Inherent Safety ◽

Oxide Fuel ◽

Fast Reactors ◽

Power Oscillation ◽

Negative Reactivity ◽

Oscillation Problem ◽

Reactivity Coefficient

The floating absorber for safety at transient (FAST) was proposed as a solution for the positive coolant temperature coefficient in sodium-cooled fast reactors (SFRs). It is designed to insert negative reactivity in the case of coolant temperature rise or coolant voiding in an inherently passive way. The use of the original FAST design showed effectiveness in protecting the reactor core during some anticipated transients without scram (ATWS) events. However, oscillation behaviors of power due to refloating of the absorber module in FAST were observed during other ATWS events. In this paper, we propose an improved FAST device (iFAST), in which a constraint is imposed on the sinking (insertion) limit of the absorber module in FAST. This provides a simple and effective solution to the power oscillation problem. Here, we focus on an oxide fuel-loaded SFR that is characterized by a more negative Doppler reactivity coefficient and higher operating temperature than the metallic-loaded SFR cores. The study is carried out for the 1000 MWth advanced burner reactor with an oxide fuel-loaded core during postulated ATWS events that are unprotected transient over power, unprotected loss of flow, and unprotected loss of the heat sink. It was found that the iFAST device has promising potentials for protecting the oxide SFR core during the various studied ATWS events.

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Analysis of Doppler reactivity of SMART reactor core for hybrid fuel configurations of UO2, MOX and (Th/U)O2 using OpenMC

Kerntechnik ◽

10.1515/kern-2020-0063 ◽

2021 ◽

Vol 86 (3) ◽

pp. 229-235

Author(s):

Y. Alzahrani ◽

K. Mehboob ◽

F. A. Abolaban ◽

H. Younis

Keyword(s):

Reactor Core ◽

Multiplication Factor ◽

The Core ◽

Mox Fuel ◽

The Monte Carlo Method ◽

Burnable Absorber ◽

Effective Multiplication Factor ◽

Reactivity Coefficient ◽

Increasing Temperature ◽

Effective Multiplication

Abstract In this study, the Doppler reactivity coefficient has been investigated for UO2, MOX, and (Th/U)O2 fuel types. The calculation has been carried out using the Monte Carlo method ( OpenMC). The effective multiplication factor keff has been evaluated for three materials with four different configurations without Integral Fuel Burnable Absorber (IFBA) rods and soluble boron. The results of MOX fuel, homogenous and heterogeneous thorium fuel configuration has been compared with the core of the reference fuel assembly (UO2). The calculation showed that the effective multiplication factor at 1 000 K was 1.26052, 1.14254, 1.22018 and 1.23771 for reference core, MOX, homogenous and heterogeneous configurations respectively. The results shows that reactivity has decreased with increasing temperature while the doppler reactivity coefficient remained negative. Moreover, the use of (Th/U)O2 homogenous and heterogeneous configuration had shown an improved response compared to the reference core at 600 K and 1 000 K. The doppler reactivity coefficient has been found as –8.98E-3 pcm/K, -0.8 655 pcmK for the homogenous and –8.854 pcm/K, -1.2253 pcm/K for the heterogeneous configuration. However, the pattern remained the same as for the reference core at other temperature points. MOX fuel has shown less response compared to the other fuel configuration because of the high resonance absorption coefficient of Plutonium. This study showed that the SMART reactor could be operated safely with investigated fuel and models.

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ANALISIS KONVEKSI ALAM TERAS REAKTOR TRIGA BERBAHAN BAKAR TIPE PELAT MENGGUNAKAN COOLOD-N2

JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA ◽

10.17146/tdm.2015.17.2.2317 ◽

2015 ◽

Vol 17 (2) ◽

pp. 67 ◽

Cited By ~ 5

Author(s):

Sudjatmi K A ◽

Endiah Puji Hastuti ◽

Surip Widodo ◽

Reinaldy Nazar

Keyword(s):

Natural Convection ◽

Fuel Element ◽

Flow Instability ◽

Reactor Core ◽

Nucleate Boiling ◽

Coolant Temperature ◽

Triga Reactor ◽

Fuel Elements ◽

Type Fuel ◽

Plate Type

ABSTRAK Analisis Konveksi Alam Teras Reaktor Triga Berbahan Bakar Tipe Pelat MENGGUNAKAN COOLOD-N2. Rencana penghentian produksi elemen bakar jenis TRIGA oleh produsen elemen bakar reaktor TRIGA, sudah seharusnya diantisipasi oleh badan pengoperasi reaktor TRIGA untuk menggantikan elemen bakar tipe silinder tersebut dengan tipe pelat yang tersedia di pasaran. Pada penelitian ini dilakukan perhitungan untuk model teras reaktor dengan spesifikasi utama menggunakan bahan bakar U3Si2Al dengan pengayaan uranium sebesar 19,75% dan tingkat muat 2,96 gU/cm3. Analisis dilakukan menggunakan program COOLOD-N2 yang tervalidasi pada konfigurasi teras TRIGA konversi berbahan bakar tipe pelat, yang tersusun atas 16 elemen bakar, 4 elemen kendali dan 1 fasilitas iradiasi yang terletak tepat di tengah teras. Hasil analisis menunjukkan bahwa dengan temperatur pendingin masuk ke teras sebesar 37oC, dan rasio faktor puncak daya radial ≤ 1,92 maka daya maksimum yang dapat dioperasikan pada moda operasi konveksi bebas adalah 600 kW. Karakteristik termohidrolika yang diperoleh antara lain adalah temperatur pendingin di sisi outlet, kelongsong dan meat masing-masing sebesar 82,39oC, 108,88oC, dan 109,02oC, pada ΔTONB (Temperature Onset of Nucleate Boiling) =7,18oC dan nilai OFIR (Onset of flow instability ratio) =1,03 Hasil yang diperoleh dari perhitungan ini diharapkan dapat dijadikan acuan untuk menentukan tingkat daya reaktor TRIGA berbahan bakar pelat. Kata kunci: TRIGA Konversi, COOLOD-N2, karakteristik termohidrolika, konveksi alam, elemen bakar tipe pelat. ABSTRACT ANALYSIS OF NATURAL CONVECTION IN TRIGA REACTOR CORE PLATE TYPES FUELED USING COOLOD-N2. Any pretensions to stop the production of TRIGA fuel elements by TRIGA reactor fuel elements manufacturer should be anticipated by the operating agency of TRIGA reactor to replace the cylinder type fuel element with plate type fuel element that available on the market. In this study, the calculation of U3Si2Al fuel with uranium enrichment of 19.75 % and a load level of 2.96 gU/cm3 was performed. Analyses were performed using the validated COOLOD - N2 program. TRIGA conversion core configurations of fuel plate type are composed of 16 fuel elements, 4 control elements and 1 irradiation facilities which are located in the middle of core. The calculation results showed that if the cooling temperature was 37°C, and the ratio of radial power peaking factor ≤ 1.92, then the maximum power that can be operated on free convection mode of operation was 600 kW. The thermalhydraulic characteristic obtained such as coolant temperature at the outlet side, cladding and meat were 82.39°C, 108.88°C and 109.02°C respectively, while the ΔTONB (Temperature Onset of Nucleate Boiling) was 7.18°C and OFIR (Onset of flow instability ratio) value was 1.03. The results are expected to be used as a reference for determining the power level of the TRIGA reactor core plate types fueled. Keywords: TRIGA Convertion, COOLOD-N2, Thermalhydraulics characteristic, natural convection, plate type fuel element.

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Use of erbium as a burnable absorber for the VVER reactor core life extension

Izvestiya Wysshikh Uchebnykh Zawedeniy Yadernaya Energetika ◽

10.26583/npe.2020.3.06 ◽

2020 ◽

Vol 2020 (3) ◽

pp. 62-71

Author(s):

Saleh Hayel Alassaf ◽

Vladimir Igorevich Savander ◽

Abyelhamd Abdelnaby Hassan

Keyword(s):

Reactor Core ◽

Vver Reactor ◽

Life Extension ◽

Burnable Absorber

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Analysis Of Doppler Reactivity Coefficient On The Typical Pwr-1000 Reactor With Mox Fuel

KnE Energy ◽

10.18502/ken.v1i1.473 ◽

2016 ◽

Vol 1 (1) ◽

Cited By ~ 1

Author(s):

Rokh Madi

Keyword(s):

Nuclear Reactor ◽

Reactor Core ◽

Nuclear Data ◽

Fuel Temperature ◽

Safety Level ◽

The Core ◽

Mox Fuel ◽

Calculation Results ◽

Reactivity Coefficient ◽

Fuel Elements

Doppler coefficient is defined as a relation between fuel temperature changes and reactivity changes in the nuclear reactor core. Doppler reactivity coefficient needs to be known because of its relation to the safety of reactor operation. This study is intended to determine the safety level of the typical PWR-1000 core by calculating the Doppler reactivity coefficient in the core with UO2 and MOX fuels. The typical PWR-1000 core is similar to the PWR AP1000 core designed by Westinghouse but without Integrated Fuel Burnable Absorber (IFBA) and Pyrex. Inside the core, there are UO2 fuel elements with 3.40 % and 4.45 % enrichment, and MOX fuel elements with 0.2 % enrichment. By its own way, the presence of Plutonium in the MOX fuel will contribute to the change in core reactivity. The calculation was conducted using MCNPX code with the ENDF/B- VII nuclear data. The reactivity change was investigated at various temperatures. The calculation results show that the core reactivity coefficient of both UO2 and MOX fuel are negative, so that the reactor is operated safely.

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Design and Calibration of a Hall Effect System for Measurement of Six-Degree-of-Freedom Motion within a Stacked Column

Sensors ◽

10.3390/s21113740 ◽

2021 ◽

Vol 21 (11) ◽

pp. 3740

Author(s):

Olafur Oddbjornsson ◽

Panos Kloukinas ◽

Tansu Gokce ◽

Kate Bourne ◽

Tony Horseman ◽

...

Keyword(s):

Hall Effect ◽

Degrees Of Freedom ◽

Reactor Core ◽

Degree Of Freedom ◽

Life Extension ◽

Scale Model ◽

Design Development ◽

Seismic Safety ◽

Six Degrees Of Freedom ◽

Six Degree Of Freedom

This paper presents the design, development and evaluation of a unique non-contact instrumentation system that can accurately measure the interface displacement between two rigid components in six degrees of freedom. The system was developed to allow measurement of the relative displacements between interfaces within a stacked column of brick-like components, with an accuracy of 0.05 mm and 0.1 degrees. The columns comprised up to 14 components, with each component being a scale model of a graphite brick within an Advanced Gas-cooled Reactor core. A set of 585 of these columns makes up the Multi Layer Array, which was designed to investigate the response of the reactor core to seismic inputs, with excitation levels up to 1 g from 0 to 100 Hz. The nature of the application required a compact and robust design capable of accurately recording fully coupled motion in all six degrees of freedom during dynamic testing. The novel design implemented 12 Hall effect sensors with a calibration procedure based on system identification techniques. The measurement uncertainty was ±0.050 mm for displacement and ±0.052 degrees for rotation, and the system can tolerate loss of data from two sensors with the uncertainly increasing to only 0.061 mm in translation and 0.088 degrees in rotation. The system has been deployed in a research programme that has enabled EDF to present seismic safety cases to the Office for Nuclear Regulation, resulting in life extension approvals for several reactors. The measurement system developed could be readily applied to other situations where the imposed level of stress at the interface causes negligible material strain, and accurate non-contact six-degree-of-freedom interface measurement is required.

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Pellet-Cladding Interaction of LMFBR Fuel Elements at Unsteady State: An Introduction to Computer Code ISUNE-5

Journal of Engineering for Power ◽

10.1115/1.3230785 ◽

1981 ◽

Vol 103 (4) ◽

pp. 627-636 ◽

Cited By ~ 1

Author(s):

B. M. Ma

Keyword(s):

Reactor Core ◽

Fatigue Cracking ◽

Computer Code ◽

Fuel Burnup ◽

Fuel Pellet ◽

Gas Release ◽

Unsteady State ◽

Fission Gas ◽

Fuel Elements ◽

Fission Gas Release

The fuel pellet-cladding interaction (PCI) of liquid-metal fast breeder reactor (LMFBR) fuel elements or fuel rods at unsteady state is analyzed and discussed based on experimental results. In the analyses, the heat generation, fuel restructuring, temperature distribution, gap conductance, irradiation swelling, irradiation creep, fuel burnup, fission gas release, fuel pellet cracking, crack healing, cladding cracking, yield failure and fracture failure of the fuel elements are taken into consideration. To improve the sintered (U,Pu)O2 fuel performance and reactor core safety at high temperature and fuel burnup, it is desirable to (a) increase and maintain the ductility of cladding material, (b) provide sufficient gap thickness and plenum space for accommodating fission gas release, (c) keep ramps-power increase rate slow and gentle, and (d) reduce the intensity and frequency of transient PCI in order to avoid intense stress fatigue cracking (SFC) and stress corrosion cracking (SCC) due to fission product compounds CsI, CdI2, Cs2Te, etc. at the inner cladding surface of the fuel elements during PCI.

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Improvement of Temperature Evaluation Model of Biological Shielding Concrete for HTTR Test Simulating LOFC With VCS Inactive

Volume 2: Plant Systems, Construction, Structures and Components; Next Generation Reactors and Advanced Reactors ◽

10.1115/icone21-16095 ◽

2013 ◽

Author(s):

Shoji Takada ◽

Shunki Yanagi ◽

Kazuhiko Iigaki ◽

Masanori Shinohara ◽

Daisuke Tochio ◽

...

Keyword(s):

Heat Transfer ◽

Natural Convection ◽

Thermal Radiation ◽

Cooling System ◽

Evaluation Model ◽

Reactor Core ◽

Reactor Power ◽

Coolant Temperature ◽

Water Cooling ◽

Reactor Outlet

HTTR is a helium gas cooled graphite-moderated HTGR with the rated power 30 MWt and the maximum reactor outlet coolant temperature 950°C. The vessel cooling system (VCS), which is composed of thermal reflector plates, cooling panel composed of fins connected between adjacent water cooling tubes, removes decay heat from reactor core by heat transfer of thermal radiation, conduction and natural convection in case of loss of forced cooling (LOFC). The metallic supports are embedded in the biological shielding concrete to support the fins of VCS. To verify the inherent safety features of HTGR, the LOFC test is planned by using HTTR with the VCS inactive from an initial reactor power of 9 MWt under the condition of LOFC while the reactor shut-down system disabled. In this test, the temperature distribution in the biological shielding concrete is prospected locally higher around the support because of thermal conduction in the support. A 2-dimensional symmetrical model was improved to simulate the heat transfer to the concrete through the VCS support in addition to the heat transfer thermal radiation and natural convection. The model simulated the water cooling tubes setting horizontally at the same pitch with actual configuration. The numerical results were verified in comparison with the measured data acquired from the test, in which the RPV was heated up to around 110 °C without nuclear heating with the VCS inactive, to show that the temperature is locally high but kept sufficiently low around the support in the concrete due to sufficient thermal conductivity to the cold temperature region.

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A Finnish District Heating Reactor: Neutronics Design and Fuel Cycle Simulations

10.1115/icone28-64347 ◽

2021 ◽

Author(s):

Jaakko Leppänen ◽

Ville Valtavirta ◽

Riku Tuominen ◽

Antti Rintala ◽

Unna Lauranto

Keyword(s):

Fuel Cycle ◽

Natural Circulation ◽

Reactor Core ◽

District Heating ◽

Research Centre ◽

Computational Framework ◽

Conceptual Design Phase ◽

Burnable Absorber ◽

Control Rods ◽

Type Fuel

Abstract The development of a small PWR for district heating applications has been started at VTT Technical Research Centre of Finland, and the pre-conceptual design phase was completed by the end of year 2020. The heating plant consists of one or multiple 50 MW reactor modules, operating on natural circulation at around 120°C temperature. This paper presents the neutronics design and fuel cycle simulations carried out using VTT’s Kraken computational framework. The reactor is operated without soluble boron, which together with low operating temperature and pressure brings certain challenges to the use of control rods and burnable absorber. The reactor core is loaded with 37 truncated AP1000-type fuel assemblies with 2.0–3.0% fuel enrichment and erbium burnable absorber. The resulting cycle length is around 900 days. The results show that the criteria set for stability, reactivity control and thermal margins are fulfilled. More importantly, it is concluded that the new Kraken framework is a viable tool for the core design task.

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Modeling the Spectral History in the Depletion of a PWR Core

18th International Conference on Nuclear Engineering: Volume 2 ◽

10.1115/icone18-29748 ◽

2010 ◽

Author(s):

I. Bilodid

Keyword(s):

Cross Sections ◽

Uranium Oxide ◽

Reactor Core ◽

Actual Distribution ◽

History Effects ◽

Fuel Assemblies ◽

Burnable Absorber ◽

Uranium Plutonium ◽

Moderator Density ◽

Local Parameters

Codes for reactor core calculations use few-group cross sections (XS) which depend on local burnup, given in terms of the energy produced per fuel mass (MWd/kgHM). However, a certain burnup value can be reached under different spectral conditions depending on moderator density and other local parameters. Neglecting these spectral effects, i.e. applying the summary-burnup value only, can cause considerable errors in the calculated power density. This paper describes a way to take into account spectral-history effects. It is shown that the respective XS correction linearly depends on the actual Pu-239 concentration. The applicability of the method was proved not only for usual uranium oxide fuel, but also for mixed uranium/plutonium oxide (MOX) and fuel assemblies with burnable absorber. The code DYN3D was extended by new subroutines which calculate the actual distribution of Pu-239 in the core and apply a spectral-history correction for the XS.

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