Coupled code calculation of rod withdrawal at power accident

Grgić, Davor; Benčik, Vesna; Šadek, Siniša

doi:10.1016/j.nucengdes.2012.10.026

Cited by 7 publications

(6 citation statements)

References 1 publication

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“…In addition, the qualification of CASMO-4/PARCS/RELAP-5 was assessed during the neutronics design of the Purdue University 200 MWe SBWR (Tinkler and Downar, 2003). Furthermore, extensive coupled codes (PARCS/RELAP) technique development and validation, regarding steady-state as well as transient scenarios, have been performed to ensure proper consideration of important thermal feedbacks (Abarca et al, 2011;Agung et al, 2013;Grgić et al, 2013;Kozlowski et al, 2004).…”

Section: Analysis Methods and Modelsmentioning

confidence: 99%

Core design optimization and analysis of the Purdue Novel Modular Reactor (NMR-50)

Odeh¹,

Yang²

2016

Annals of Nuclear Energy

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A BWR-based SMR called the Novel Modular Reactor (NMR-50) is being developed at Purdue University. NMR takes the advantages of the two-phase flow driving head, which allows a much smaller and simpler reactor pressure vessel (RPV) compared to the integral PWRs. In this study, through a systematic step-wise optimization approach including a simulated annealing based optimization method, an optimum core design that meets a 10-year cycle length with a minimum fuel cost while satisfying safety related criteria was derived and analyzed. The lattice code CASMO-4, the whole core analysis code PARCS and the thermal-hydraulics code RELAP5 were used to perform calculations from pin cell up to whole core depletion calculations. The NMR-50 optimized core design is able to achieve a 10.2 year cycle length with an average fuel enrichment of 4.61 wt% of 235 U in a 1010 lattice fuel assembly. The minimum critical power ratio (MCPR) and the maximum fuel linear power density (MFLPD) during the cycle are 1.99 and 18.25 kW/m, respectively, providing large margins to thermal design constraints. The NMR-50 control system design is able to provide a sufficient cold shutdown margin of 1.7 %. With its small reactor core size, large negative void coefficient, and low operating thermal neutron flux, an enhanced xenon stability characteristic is possible. Peak fast neutron fluence of 21 2 8.8 10 n cm  was below the industry standard limit, which from extensive plant data records, should not be a major concern to channel distortions from a radiation damage point of view.

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Section: Analysis Methods and Modelsmentioning

confidence: 99%

Core design optimization and analysis of the Purdue Novel Modular Reactor (NMR-50)

Odeh¹,

Yang²

2016

Annals of Nuclear Energy

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“…All calculations were carried out by the adiabatic method (AM) and improved quasistatic method (IQM), which are adopted in the NODAL3 code [5]. The core configuration is modeled by 3-dimension of (X-Y-Z) geometry with 2 × 2 × 1 nodes for one assembly per layer.…”

Section: Calculation Methodsmentioning

confidence: 99%

“…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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“…Code coupling is a standard methodology used to describe transients having spatial reactivity dependence in the core and thermal-hydraulic influence from rest of the system [7]. The coupled code RELAP5/mod3.3-PARCS v2.5 (R5PA) has been developed at FER.…”

Section: Coupled Relap5 and Parcs Programmentioning

confidence: 99%

NPP Krško Post-UFC Transient Response during MSLB

Vlahović¹,

Grgić²,

Benčik³

2022

joe

Self Cite

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UpFlow Conversion (UFC) was implemented in NPP Krško during the last outage in order to reduce the pressure differential across baffle plates and the possibility of the fuel damage caused by flow induced vibration. The paper describes the coupled code calculation (RELAP5 and PARCS) of MSLB accident at power for pre and post-UFC configuration of reactor vessel. In the calculation, the split model of the reactor vessel was used to better describe asymmetric conditions in loops. It has been demonstrated that the basic parameters (pressure, temperatures) stayed unchanged and there was little change in the flow rates except in baffle-barrel region of the vessel where both flow direction and amount of flow were changed.

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Coupled code calculation of rod withdrawal at power accident

Cited by 7 publications

References 1 publication

Core design optimization and analysis of the Purdue Novel Modular Reactor (NMR-50)

Core design optimization and analysis of the Purdue Novel Modular Reactor (NMR-50)

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

NPP Krško Post-UFC Transient Response during MSLB

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