Steady state and transient thermalhydraulic analysis of PHWR using COBRA-3C/RERTR

Nuclear cross sections that determine core multiplication strongly depend on core temperature (e.g., the Doppler, moderator density effects etc). On the other hand, since this heat is generated by the neutron flux in the reactor core, the temperature distribution in the core will depend heavily on i...

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Main Authors: Hussain, A., Abolaban, F., Khubaib, S.M., Mubin, S., Ahmed, I.
Format: Article
Institution: Universiti Teknologi Petronas
Record Id / ISBN-0: utp-eprints.26007 /
Published: Springer International Publishing 2015
Online Access: https://www.scopus.com/inward/record.uri?eid=2-s2.0-85033361778&partnerID=40&md5=3b5766da0f30acb9dec232ec2c479b68
http://eprints.utp.edu.my/26007/
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spelling utp-eprints.260072021-08-30T08:49:43Z Steady state and transient thermalhydraulic analysis of PHWR using COBRA-3C/RERTR Hussain, A. Abolaban, F. Khubaib, S.M. Mubin, S. Ahmed, I. Nuclear cross sections that determine core multiplication strongly depend on core temperature (e.g., the Doppler, moderator density effects etc). On the other hand, since this heat is generated by the neutron flux in the reactor core, the temperature distribution in the core will depend heavily on its neutronic behavior. Fuel centerline temperature could be the limiting constraint on reactor power because of the concern for fuel melting. Likewise, high clad temperature is also a possible limiting factor on reactor power because of the potential degradation of clad material or on-set of critical heat flux phenomenon. An assessment of the steady state and transient thermal hydraulic capabilities of the computer code COBRA 3C/RERTR was made using model for a PHWRs reactor core. The temperature distributions determined for fuel, clad and coolant are compared with analytical results and with the results quoted in safety report. It was found that when the code was run for full power at reduced flow of 70 the bulk coolant temperature remained below the saturation temperature, so there is an adequate design margin is available for safety related scenarios. © Shiraz University. Springer International Publishing 2015 Article NonPeerReviewed https://www.scopus.com/inward/record.uri?eid=2-s2.0-85033361778&partnerID=40&md5=3b5766da0f30acb9dec232ec2c479b68 Hussain, A. and Abolaban, F. and Khubaib, S.M. and Mubin, S. and Ahmed, I. (2015) Steady state and transient thermalhydraulic analysis of PHWR using COBRA-3C/RERTR. Iranian Journal of Science and Technology - Transactions of Mechanical Engineering, 39 . pp. 233-242. http://eprints.utp.edu.my/26007/
institution Universiti Teknologi Petronas
collection UTP Institutional Repository
description Nuclear cross sections that determine core multiplication strongly depend on core temperature (e.g., the Doppler, moderator density effects etc). On the other hand, since this heat is generated by the neutron flux in the reactor core, the temperature distribution in the core will depend heavily on its neutronic behavior. Fuel centerline temperature could be the limiting constraint on reactor power because of the concern for fuel melting. Likewise, high clad temperature is also a possible limiting factor on reactor power because of the potential degradation of clad material or on-set of critical heat flux phenomenon. An assessment of the steady state and transient thermal hydraulic capabilities of the computer code COBRA 3C/RERTR was made using model for a PHWRs reactor core. The temperature distributions determined for fuel, clad and coolant are compared with analytical results and with the results quoted in safety report. It was found that when the code was run for full power at reduced flow of 70 the bulk coolant temperature remained below the saturation temperature, so there is an adequate design margin is available for safety related scenarios. © Shiraz University.
format Article
author Hussain, A.
Abolaban, F.
Khubaib, S.M.
Mubin, S.
Ahmed, I.
spellingShingle Hussain, A.
Abolaban, F.
Khubaib, S.M.
Mubin, S.
Ahmed, I.
Steady state and transient thermalhydraulic analysis of PHWR using COBRA-3C/RERTR
author_sort Hussain, A.
title Steady state and transient thermalhydraulic analysis of PHWR using COBRA-3C/RERTR
title_short Steady state and transient thermalhydraulic analysis of PHWR using COBRA-3C/RERTR
title_full Steady state and transient thermalhydraulic analysis of PHWR using COBRA-3C/RERTR
title_fullStr Steady state and transient thermalhydraulic analysis of PHWR using COBRA-3C/RERTR
title_full_unstemmed Steady state and transient thermalhydraulic analysis of PHWR using COBRA-3C/RERTR
title_sort steady state and transient thermalhydraulic analysis of phwr using cobra-3c/rertr
publisher Springer International Publishing
publishDate 2015
url https://www.scopus.com/inward/record.uri?eid=2-s2.0-85033361778&partnerID=40&md5=3b5766da0f30acb9dec232ec2c479b68
http://eprints.utp.edu.my/26007/
_version_ 1741197068253265920
score 11.62408

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