Thermal-hydraulic analysis of VVER-1000 residual heat removal system using RELAP5 code, an evaluation at the boundary of reactor repair mode

Tabadar, Z.; Jabbari, Masoud; Khaleghi, M.; Hashemi-Tilehnoee, M.

doi:10.1016/j.aej.2017.03.044

Cited by 17 publications

(5 citation statements)

References 14 publications

(15 reference statements)

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“…The results showed that the inherent safety of secondary PRHRS is better, and the area of heat transfer and the vertical distance of cold and heat sources are the main factors affecting the system. Tabadar et al 13 used RELAP5 to evaluate the PRHRS of VVER‐1000, and the calculation results were consistent with experimental data and safety analysis reports. It was recommended that RELAP5 could be used to analyze VVER‐1000 during reactor cooling.…”

Section: Introductionmentioning

confidence: 55%

Design and optimization of passive residual heat removal system for lead‐bismuth reactorSVBR‐100

Cheng

Zhang

et al. 2020

Int J Energy Res

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Summary The passive residual heat removal system (PRHRS) is of great significance to ensure the safety of small modular reactors. In this work, a PRHRS based on the reactor pressure vessel was designed and optimized for the lead‐bismuth reactor SVBR‐100 and an one‐dimensional transient analysis code transient analysis codes for LBE reactor was developed as well. The uncertainty analysis was carried out to identify and optimize the key parameters affecting PRHRS, based on the one‐dimensional physical fields obtained from the primary loop and evaporator. The results of transient calculation showed that the designed PRHRS can effectively remove the heat of the core under the accident of unprotected loss of flow and the station blackout (SBO). The results of sensitivity analysis showed that the maximum temperature of lead‐bismuth eutectic (LBE) was strongly positively correlated with heat capacity and thermal conductivity of the cladding, and was moderately positively correlated with the kinetic coefficient of the pump; The medium negative correlation factor of peak pellet temperature (PPT), peak cladding temperature (PCT), and maximum LBE temperature was the diameter of the sleeve of PRHRS, and therefore, increasing the diameter of the sleeve can improve the performance of the system. The results of uncertainty quantification showed that the output uncertainties of PPT and PCT were larger than those inputs. The uncertainty of LBE temperature was very low, and all thermal parameters in SBO were within safe limits. Finally, the optimized parameters of PRHRS were obtained. This work can provide a reference for the development of designing PRHRS and relevant transient calculation codes for small modular LBE reactors.

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Section: Introductionmentioning

confidence: 55%

Design and optimization of passive residual heat removal system for lead‐bismuth reactorSVBR‐100

Cheng

Zhang

et al. 2020

Int J Energy Res

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“…For VVER-1000, s = 12.75 mm, and d = 9.1 mm [33]. k 3 accounts for the heated length effect and is analogous to Groeneveld's K 4 .…”

Section: Bobkov Look-up Tablementioning

confidence: 99%

“…The degree to which the assemblies used for constructing the LUT were "unbalanced" varied from 0.42 to 0.93 [8]; the value 0.93 corresponding to a bundle containing 37 rods. The number of fuel rods in a VVER-1000 assembly is 311 [33], so we can safely assume that k 5 = 1.…”

Section: Bobkov Look-up Tablementioning

confidence: 99%

A Methodology for CHF Prediction in VVER Rod Bundles

AbdulHameed,

Shaaban,

Gamal

et al. 2022

Preprint

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The critical heat flux (CHF) is an important thermal-hydraulic parameter that must be determined during the design of water-cooled reactors to ensure safety of operation. In this paper, a methodology is proposed to predict the CHF in VVER 1 rod bundles. A whole-reactor model of VVER-1000 is implemented in RELAP-SCDAPSIM, and, based on the bundle-average approach, the parameters of the hot channel are used to assess four CHF predictors (i.e., the 2006 Groeneveld look-up table (LUT), and the Bowring, W-3, Biasi, and OKB-Gidropress correlations) upon comparison with the 2011 Bobkov LUT. A brief critique is given of the correction factors for both the Groeneveld and Bobkov LUTs. The effects of channel diameter and non-uniform axial heat flux on the predicted values of CHF in subcooled conditions are also studied. The performance and time complexity of the standard and Bourke trilinear interpolation algorithms are assessed for use with LUTs in safety codes. It has been concluded that the non-uniform axial heat flux does not affect the CHF in subcooled conditions and that the Groeneveld diameter factor exponent gives better predictability of the CHF in VVER rod bundles over the Tanase and Wong exponent correlations. Additionally, the Groeneveld LUT has been found to nearly reproduces the Bobkov LUT CHF values when used in conjunction with the Bobkov heated length factor. Values of MDNBR equal to 1.83 and 1.96 can be used as thermal limits at 12% overpower when the Bowring and W-3 correlations, respectively, are used in subcooled conditions. The standard and Bourke algorithms nearly exhibit the same average CPU time. The Bourke algorithm, however, is less affected by noise. Further research is needed to ensure the smoothness of the Bobkov LUT and to derive reactor type-dependent correction factor formulas for use with the Groeneveld LUT.

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“…For the VVER series, the VVER-1000/V412, V428M, and V528 are equipped with passive residual heat removal system (PRHRS) consisting of 4 trains with 33% capacity each, in which each train is connected into 1 horizontal steam generator. The steam line of steam generator has a bypass line into the finned tube air heat exchangers, which are used to reject core heat to the outside atmosphere [18,19,20]. There are 4 HXs for each train inside a chimney-like structure along the outer surface of containment up to the top of the containment to enable a natural circulation of air as shown in Figure 5.…”

Section: Engineered Safety Features In Case Of Sbomentioning

confidence: 99%

Preliminary Assessment of Engineered Safety Features Against Station Blackout in Selected PWR Models

Ekariansyah

Widodo²,

Susyadi³

et al. 2021

Tri Dasa Mega

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The 2011 Fukushima accident did not prevent countries to construct new nuclear power plants (NPPs) as part of the electricity generation system. Based on the IAEA database, there are a total of 44 units of PWR type NPPs whose constructions are started after 2011. To assess the technology of engineered safety features (ESFs) of the newly constructed PWRs, a study has been conducted as described in this paper, especially in facing the station blackout (SBO) event. It is expected from this study that there are a number of PWR models that can be considered to be constructed in Indonesia from the year of 2020. The scope of the study is PWRs with a limited capacity from 900 to 1100 MWe constructed and operated after 2011 and small-modular type of reactors (SMRs) with the status of at least under licensing. Based on the ESFs design assessment, the passive core decay heat removal has been applied in the most PWR models, which is typically using steam condensing inside heat exchanger within a water tank or by air cooling. From the selected PWR models, the CPR-1000, HPR-1000, AP-1000, and VVER-1000, 1200, 1300 series have the capability to remove the core decay heat passively. The most innovative passive RHR of AP-1000 and the longest passive RHR time period using air cooling in several VVER models are preferred. From the selected SMR designs, the NuScale design and RITM-200 possess more advantages compared to the ACP-100, CAREM-25, and SMART. NuScale represents the model with full-power natural circulation and RITM-200 with forced circulation. NuScale has the longest time period for passive RHR as claimed by the vendor, however the design is still under licensing process. The RITM-200 reactor has a combination of passive air and water-cooling of the heat exchanger and is already under construction.

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Thermal-hydraulic analysis of VVER-1000 residual heat removal system using RELAP5 code, an evaluation at the boundary of reactor repair mode

Cited by 17 publications

References 14 publications

Design and optimization of passive residual heat removal system for lead‐bismuth reactorSVBR‐100

Design and optimization of passive residual heat removal system for lead‐bismuth reactorSVBR‐100

A Methodology for CHF Prediction in VVER Rod Bundles

Preliminary Assessment of Engineered Safety Features Against Station Blackout in Selected PWR Models

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