scholarly journals A SIMPLIFIED SMAHTR BENCHMARK PROBLEM SET

2021 ◽  
Vol 247 ◽  
pp. 10023
Author(s):  
K. Lisa Reed ◽  
Farzad Rahnema
Keyword(s):  
Neutron Transport ◽  
Benchmark Problem ◽  
Control Rod ◽  
Burnable Poison ◽  
Continuous Energy ◽  
Small Modular Reactor ◽  
High Temperature Reactor ◽  
And Control ◽  
Control Rods ◽  
Full Core

The Small, Modular Advanced High Temperature Reactor (SmAHTR) is a preconceptual design for a fluoride salt-cooled small modular reactor (SMR) [1]. In this paper, a stylized 2D benchmark problem set has been created based on SmAHTR. Certain gaps and considerations in burnable poison and control rod content were unspecified/undetermined in the preconceptual design, but those gaps were filled for the stylized problem set. With those features, this problem set could then be used for benchmarking neutron transport methods as well as its low order methods in 2D single assembly and full core configurations. For this benchmark set, continuous energy Monte Carlo calculations were performed. Those calculations provided keff values of 0.9459 (± 11 pcm) and 1.1436 (± 12 pcm) in the full core configuration with all the control rods fully inserted and withdrawn, respectively. The single assembly calculations yielded an eigenvalue, kinf, of 0.9987 (± 15 pcm) and 1.2117 (± 15 pcm) with all of the control rods either inserted or removed, respectively. In the full core configuration, the worth of all the control rods and burnable poison particles were determined to be 197.6 (± 0.16) mk and 311.6 (± 0.23) mk, respectively. The corresponding results in the single assembly configurations are 213 (± 0.21) mk and 337.4 (± 0.20) mk, respectively. A near-critical configuration was also determined for the reactor by inserting control rods in some assemblies, thus providing a case with a keff value of 0.9909 (± 12 pcm).

Control Strategy of Soluble Boron-Free Operation for a Small PWR

2017 ◽  
Author(s):  
Guangwen Bi ◽  
Chuntao Tang ◽  
Bo Yang
Keyword(s):  
Control Strategy ◽  
Power Distribution ◽  
Short Term ◽  
Control Rod ◽  
Burnable Poison ◽  
Power Operation ◽  
And Control ◽  
Power Distribution Control ◽  
Selection Of

Elimination of soluble boron will be a challenge to reactor operation for PWR. This paper is to promote a control strategy of soluble boron-free operation for a small PWR, through selection of burnable poison (BP), BP loading and control rod loading, based on the reactivity balance and manage requirement. The analysis for on-power operation and shutdown condition indicated that this strategy could be suitable for long-term and short-term reactivity and power distribution control for soluble boron-free operation.

Impact of control rod insertion during burnup on PWR fuel assembly isotopic composition

Kerntechnik ◽  
2021 ◽  
Vol 86 (2) ◽  
pp. 173-181
Author(s):  
R. M. Refeat ◽  
H. K. Louis
Keyword(s):  
Safety Analysis ◽  
Second Phase ◽  
Spent Fuel ◽  
Reactor Operation ◽  
Control Rod ◽  
Two Phases ◽  
And Control ◽  
Criticality Analysis ◽  
Control Rods ◽  
Criticality Safety

Abstract Criticality analysis of spent fuel assumes that the fuel material is unburned which means that it is in its most reactive condition. In fact, this is not the real situation for fuel as it is burned during reactor operation causing reduction in the reactivity. Considering the reduction in reactivity during spent fuel calculations is the Burn-up Credit concept (BUC). In addition, the control rods radial and axial positions have an effect on the reactivity which can be considered in the criticality safety analysis. This paper studies the effect of burnup and control rods (CRs) movement on reactivity and isotopes inventory. Calculations are carried out in two phases, first kinf is calculated for different burnup profiles with control rods are either fully withdrawn or fully inserted. In the second phase keff is calculated for different control rods insertion levels. For both phases, burnup calculations are performed for a UO2 assembly then multiplication factor calculations of burned UO2 assemblies in cold state are done. The burnup calculations are performed using MCNP6 code and ENDF/B-VII library for different burnup levels up to 45 GWd/tU. The results obtained can be taken in consideration in criticality safety analysis performed for the spent fuel to improve the economic efficiency for manufacture, storage and transportation of fissile materials.

scholarly journals Development and Application of MCNP5 and KENO-VI Monte Carlo Models for the Atucha-2 PHWR Analysis

2011 ◽  
Vol 2011 ◽  
pp. 1-7 ◽  
Author(s):  
M. Pecchia ◽  
C. Parisi ◽  
F. D'Auria ◽  
O. Mazzantini
Keyword(s):  
Monte Carlo ◽  
Cross Sections ◽  
3D Model ◽  
Three Dimensional ◽  
Control Rod ◽  
Geometrical Complexity ◽  
Reactor Criticality ◽  
Reactor Cell ◽  
And Control ◽  
Control Rods

The geometrical complexity and the peculiarities of Atucha-2 PHWR require the adoption of advanced Monte Carlo codes for performing realistic neutronic simulations. Core models of Atucha-2 PHWR were developed using both MCNP5 and KENO-VI codes. The developed models were applied for calculating reactor criticality states at beginning of life, reactor cell constants, and control rods volumes. The last two applications were relevant for performing successive three dimensional neutron kinetic analyses since it was necessary to correctly evaluate the effect of each oblique control rod in each cell discretizing the reactor. These corrective factors were then applied to the cell cross sections calculated by the two-dimensional deterministic lattice physics code HELIOS. These results were implemented in the RELAP-3D model to perform safety analyses for the licensing process.

scholarly journals Benchmarking the new ENDF/B-VIII.0 nuclear data library for OECD/NEA medium 1000 MWth sodium-cooled fast reactor

2020 ◽  
Vol 239 ◽  
pp. 22006
Author(s):  
Donny Hartanto ◽  
Bassam Khuwaileh ◽  
Peng Hong Liem
Keyword(s):  
Monte Carlo ◽  
Fast Reactor ◽  
Nuclear Data ◽  
Control Rod ◽  
Continuous Energy ◽  
Nuclear Data Library ◽  
And Control ◽  
Benchmark Evaluation ◽  
Data Library

This paper presents the benchmark evaluation of the new ENDF/B-VIII.0 nuclear library for the OECD/NEA Medium 1000 MWth Sodium-cooled Fast Reactor (SFR). There are 2 SFR cores: metallic fueled (MET-1000) and oxide fueled (MOX-1000). The continuous-energy Monte Carlo Serpent2 code was used as the calculation tool. Various nuclear libraries such as ENDF/B-VII.1 and JENDL-4.0 were included to be compared with the newest ENDF/B-VIII.0. The evaluated parameters are k,βeff, sodium void reactivity (∆ρNa), Doppler constant (∆ρDoppler), and control rod worth (∆ρCR).

scholarly journals TRANSIENT CALCULATIONS OF SPERT III EXPERIMENTS

2021 ◽  
Vol 247 ◽  
pp. 07017
Author(s):  
Andreas Pautz ◽  
Winfried Zwermann
Keyword(s):  
Cross Sections ◽  
Reference Model ◽  
Neutron Transport ◽  
Control Rod ◽  
Combined Application ◽  
Experimental Values ◽  
Core Characteristics ◽  
Control Rods ◽  
Static Core ◽  
Good Agreement

Cold-startup and hot-standby reactivity accident tests conducted at the SPERT III E-core research reactor are analysed with the coupled neutron-kinetic/thermal-hydraulic code system DYN3D-ATHLET. Homogenised 2-group cross sections for DYN3D are thereby generated with the Monte Carlo neutron transport code Serpent 2 in combination with the ENDF/B-VII.1 cross section library. Results in terms of maximum power, energy release, and reactivity compensation are in good agreement with the experimental values. The time-dependent contributions to the reactivity feedback are investigated for both a cold-startup test and a hot-standby test. These findings prove the suitability of the combined application of the simulation codes to predict the reactor dynamic behaviour in the event of prompt-critical and super-prompt critical transients even for small reactor cores. Furthermore, static core characteristics of the SPERT III E-core reactor at cold-startup condition are analysed with using a static DYN3D model, a detailed Serpent reference model, and a simplified Serpent model consistent with the DYN3D model. The critical control rod position and the excess reactivities of both the control rods and the transient rod obtained with the Serpent reference model are consistent with the experimental values. For the same parameters, the DYN3D model is in good agreement with the Serpent simplified model.

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