REA 3D-dynamic analysis in Almaraz NPP with RELAP5/PARCS v2.7 and SIMTAB cross-sections tables

Barrachina, T.; Garcia-Fenoll, Marina; Ánchel, F.; Miró, R.; Verdú, G.; Pereira, Cláubia; Silva, C. da; Ortego, Alberto; Martínez-Murillo, J.C.

doi:10.1016/j.pnucene.2011.07.012

Cited by 11 publications

(4 citation statements)

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“…Therefore, every fuel channel had to be simulated in RSM4 to obtain accurate predictions in the entire core. Channel grouping is not used in the thermalhydraulic simulations [31].…”

Section: Parcs/rsm4 Coupling For Transient Simulationmentioning

confidence: 99%

Analysis of Control Rod Drop Accidents for the Canadian SCWR Using Coupled 3-Dimensional Neutron Kinetics and Thermal Hydraulics

Salaun

Novog

2018

Science and Technology of Nuclear Installations

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The Canadian Supercritical Water-cooled Reactor (SCWR), a GEN IV reactor design, is a hybrid design of the well-established CANDU™ and Boiling Water Reactor with water above its thermodynamic critical point. Given the batch fueled design, control rods are used to manage the reactivity throughout the fuel cycle. This paper examines the consequences of a control rod drop accident (CRDA) for the Canadian SCWR. The asymmetry generated by the dropped rod requires an accurate 3-dimensional neutron kinetics calculation coupled to a detailed thermal-hydraulic model. Before simulating the CRDAs, the proper implementation of the 3D reactivity feedback was verified and various sensitivity studies were performed. This work demonstrates that the proposed safety systems for the SCWR core are capable of terminating the CRDA sequence prior to exceeding maximum sheath and centerline temperatures. In one instance involving a rod on the periphery of the core, the proposed trip setpoint (115% FP) was not exceeded and a new steady state was reached. Therefore it is recommended that the design also include provisions for a high-log rate and/or local Neutron Overpower Protection (NOP) trips, similar to existing CANDU designs such that reactor shutdown can be assured for such spatial anomalies.

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“…Therefore, every fuel channel had to be simulated in RSM4 to obtain accurate predictions in the entire core. Channel grouping is not used in the thermalhydraulic simulations [31].…”

Section: Parcs/rsm4 Coupling For Transient Simulationmentioning

confidence: 99%

Analysis of Control Rod Drop Accidents for the Canadian SCWR Using Coupled 3-Dimensional Neutron Kinetics and Thermal Hydraulics

Salaun

Novog

2018

Science and Technology of Nuclear Installations

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“…Some previous research works showed that the analytical tool for analyzing the transient behaviour of a REA is the coupled neutronic thermal-hydraulic codes [1][2][3][4][5][6][7]. BATAN, as a research institute, has developed a coupled neutronic thermal-hydraulic (T/H) code, NODAL3, based on the nodal multi-group neutron diffusion method with 3-dimensional core geometry using polynomial nodal method [8].…”

Section: Introductionmentioning

confidence: 99%

Validation of Full Core Geometry Model of the Nodal3 Code in the PWR Transient Benchmark Problems

Sembiring

Pinem

Liem³

2015

Tri Dasa Mega

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VALIDATION OF FULL CORE GEOMETRY MODEL OF THE NODAL3 CODE IN THE PWR TRANSIENT BENCHMARK PROBLEMS.The coupled neutronic and thermal-hydraulic (T/H) code, NODAL3 code, has been validated in some PWR static benchmark and the NEACRP PWR transient benchmark cases. However, the NODAL3 code have not yet validated in the transient benchmark cases of a control rod assembly (CR) ejection at peripheral core using a full core geometry model, the C1 and C2 cases. By this research work, the accuracy of the NODAL3 code for one CR ejection or the unsymmetrical group of CRs ejection case can be validated. The calculations by the NODAL3 code have been carried out by the adiabatic method (AM) and the improved quasistatic method (IQS). All calculated transient parameters by the NODAL3 code were compared with the reference results by the PANTHER code. The maximum relative difference of 16% occurs in the calculated time of power maximum parameter by using the IQS method, while the relative difference of the AM method is 4% for C2 case. All calculation results by the NODAL3 code shows there is no systematic difference, it means the neutronic and T/H modules are adopted in the code are considered correct. Therefore, all calculation results by using the NODAL3 code are very good agreement with the reference results.Keywords: nodal method, coupled neutronic and thermal-hydraulic code, PWR, transient case, control rod ejection. ABSTRAK VALIDASI MODEL GEOMETRI TERAS PENUH PAKET PROGRAM NODAL3 DALAM PROBLEM BENCHMARK GAYUT WAKTU PWR. Paket program kopel neutronik dan termohidraulika (T/H), NODAL3, telah divalidasi dengan beberapa kasus benchmark statis PWR dan kasus benchmark gayut waktu PWR NEACRP. Akan tetapi, paket program NODAL3 belum divalidasi dalam kasus benchmark gayut waktu akibat penarikan sebuah perangkat batang kendali (CR) di tepi teras menggunakan model geometri teras penuh, yaitu kasus C1 dan C2. Dengan penelitian ini, akurasi paket program NODAL3 untuk kasus penarikan sebuah CR atau sekelompok CR yang tidak-simetris dapat divalidasi. Perhitungan paket program NODAL3 dilakukan dengan metode adiabatic (AM) dan improved quasistatic (IQS

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“…On the other hand, SIMTAB (Barrachina et al, 2011) is a well and reliable validated method. Its robust source code and the experience facilitates the use and helps to track possible programming errors when adapting the SIMTAB methodology to new and more detailed core designs.…”

Section: Simtab Methodologymentioning

confidence: 99%

Development and validation of a multi-scale and multi-physics methodology for the safety analysis of fast transients in Light Water Reactors

García-Bermejo¹

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Deferred neutron precursor fraction of the k-group (non-dimensional) Σ , Macroscopic transport cross section of energy group (cm -1 ) Σ , ⟶ ′ Macroscopic scattering cross section of energy group to ′ (cm -1 ) Σ , Macroscopic fission cross section of energy group (cm -1 ) Σ , Macroscopic absorption cross section of energy group (cm -1 ) Volumetric force tensor applied to the k-phase of a multiphase fluid (N/m 3 ) Deferred neutron decay constant of the precursor group (s -1 ) Dynamic Viscosity (Pa-s) Average Neutrons Emitted by fission (non-dimensional) Density (kg/m 3 ) Density of the -phase of a multiphase fluid (kg/m 3 ) Neutron Flux for energy group (neutrons/cm 2 ) Fission spectrum of the neutrons of the energy group Cross section area (m 2 ) ⃗ Area Vector (m 2 ) Drag coefficient (non-dimensional) Specific Heat Capacity of the fuel pellet (J/kg-K) Concentration of delayed neutron precursors of the -group (neutrons/cm 3 ) Neutron diffusion constant of the energy group (cm -1 ) Conservation of energy Gravity acceleration (m/s 2 ) ℎ Fuel Pellet Enthalpy (J/kg). Velocity Scalar Component parallel to axis (m/s) V Volume (m 3 ) Velocity Scalar Component parallel to axis (m/s) Mean neutron velocity of the neutrons in the energy group (cm/s) Work per volume unit (J/m 3 ) Velocity Scalar Component parallel to axis (m/s)To all the workers of the Scientific Research Field, that with his compromise struggle to make of this profession a dignified one.

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REA 3D-dynamic analysis in Almaraz NPP with RELAP5/PARCS v2.7 and SIMTAB cross-sections tables

Cited by 11 publications

References 0 publications

Analysis of Control Rod Drop Accidents for the Canadian SCWR Using Coupled 3-Dimensional Neutron Kinetics and Thermal Hydraulics

Analysis of Control Rod Drop Accidents for the Canadian SCWR Using Coupled 3-Dimensional Neutron Kinetics and Thermal Hydraulics

Validation of Full Core Geometry Model of the Nodal3 Code in the PWR Transient Benchmark Problems

Development and validation of a multi-scale and multi-physics methodology for the safety analysis of fast transients in Light Water Reactors

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