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Dissertations / Theses on the topic 'Reactor system code'

Author: Grafiati

Published: 4 June 2021

Last updated: 8 February 2022

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1

Breijder, Paul. "Analysis of Advanced Fuel Behaviour during Loss of Coolant Accident in Swedish Boiling Water Reactor." Thesis, KTH, Kärnkraftsäkerhet, 2011. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-44484.

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In accident analysis regarding nuclear power plants, it is very common to use thermal hydraulic system codes, such as TRACE, developed by U.S. NRC. In the case of licensing a power plant, this is one of the necessities. TRACE is a relatively new thermal hydraulic system code and a lot of knowledge is needed to implement it in a correct way, especially in accident analysis, where it is a requirement that the rules and statements in Appendix-K, dealing with criteria for ECCS-models, are modelled. In this thesis an improved model of a Swedish Boiling Water Reactor within TRACE is realized and tes

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2

Pegonen, Reijo. "Development of an Improved Thermal-Hydraulic Modeling of the Jules Horowitz Reactor." Doctoral thesis, KTH, Reaktorteknologi, 2017. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-197712.

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The newest European high performance material testing reactor, the Jules Horowitz Reactor, is under construction at CEA Cadarache research center in France. The reactor will support existing and future nuclear reactor technologies, with the first criticality expected at the end of this decade. The current/reference CEA methodology for simulating the thermalhydraulic behavior of the reactor gives reliable results. The CATHARE2 code simulates the full reactor circuit with a simplified approach for the core. The results of this model are used as boundary conditions in a three-dimensional FLICA4 c

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3

Phung, Viet-Anh. "Input Calibration, Code Validation and Surrogate Model Development for Analysis of Two-phase Circulation Instability and Core Relocation Phenomena." Doctoral thesis, KTH, Kärnkraftssäkerhet, 2017. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-202957.

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Code validation and uncertainty quantification are important tasks in nuclear reactor safety analysis. Code users have to deal with large number of uncertain parameters, complex multi-physics, multi-dimensional and multi-scale phenomena. In order to make results of analysis more robust, it is important to develop and employ procedures for guiding user choices in quantification of the uncertainties. The work aims to further develop approaches and procedures for system analysis code validation and application to practical problems of safety analysis. The work is divided into two parts. The f

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4

Grundmann, Ulrich, Ulrich Rohde, Siegfried Mittag, and Sören Kliem. "DYN3D version 3.2 - code for calculation of transients in light water reactors (LWR) with hexagonal or quadratic fuel elements - description of models and methods -." Forschungszentrum Dresden, 2010. http://nbn-resolving.de/urn:nbn:de:bsz:d120-qucosa-28604.

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DYN3D is an best estimate advanced code for the three-dimensional simulation of steady-states and transients in light water reactor cores with quadratic and hexagonal fuel assemblies. Burnup and poison-dynamic calculations can be performed. For the investigation of wide range transients, DYN3D is coupled with system codes as ATHLET and RELAP5. The neutron kinetic model is based on the solution of the three-dimensional two-group neutron diffusion equation by nodal expansion methods. The thermal-hydraulics comprises a one- or two-phase coolant flow model on the basis of four differential balanc

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5

Grundmann, Ulrich, Ulrich Rohde, Siegfried Mittag, and Sören Kliem. "DYN3D version 3.2 - code for calculation of transients in light water reactors (LWR) with hexagonal or quadratic fuel elements - description of models and methods -." Forschungszentrum Rossendorf, 2005. https://hzdr.qucosa.de/id/qucosa%3A21687.

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DYN3D is an best estimate advanced code for the three-dimensional simulation of steady-states and transients in light water reactor cores with quadratic and hexagonal fuel assemblies. Burnup and poison-dynamic calculations can be performed. For the investigation of wide range transients, DYN3D is coupled with system codes as ATHLET and RELAP5. The neutron kinetic model is based on the solution of the three-dimensional two-group neutron diffusion equation by nodal expansion methods. The thermal-hydraulics comprises a one- or two-phase coolant flow model on the basis of four differential balanc

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6

He, Xun [Verfasser], Rafael [Akademischer Betreuer] [Gutachter] Macián-Juan, and Konrad [Gutachter] Czerski. "Validation of the TRACE Code for the System Dynamic Simulations of the Molten Salt Reactor Experiment and the Preliminary Study on the Dual Fluid Molten Salt Reactor / Xun He. Betreuer: Rafael Macián-Juan. Gutachter: Konrad Czerski ; Rafael Macián-Juan." München : Universitätsbibliothek der TU München, 2016. http://d-nb.info/110564667X/34.

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7

SILVEIRA, RENATO C. da. "Avaliacao da estabilidade estrutural de contencoes metalicas de centrais nucleares." reponame:Repositório Institucional do IPEN, 2000. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10795.

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8

Al-Ani, Jonathan. "Development of a Nordic BWR plant model in APROS and design of a power controller using the control rods." Thesis, KTH, Fysik, 2021. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-289560.

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In this master thesis an input-model of a Nordic BWR power plant has been developed in APROS. The plant model contains key systems and major thermohydraulic components of the steam cycle, including I&C systems (i.e. power, pressure, level and flow controls). The plant model is primarily designed for balance of plant studies at discrete power levels. The input-model of the power plant focuses especially on the steam cycle which is crucial for analysing water and steam behaviour and its influence on the reactor power. At the current stage, the model primarily handles steady-state conditions

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9

MELLO, JOSÉ ROBERTO de. "Regulamentação do sistema elétrico do reator IEA-R1." reponame:Repositório Institucional do IPEN, 2016. http://repositorio.ipen.br:8080/xmlui/handle/123456789/26928.

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Submitted by Marco Antonio Oliveira da Silva (maosilva@ipen.br) on 2016-12-21T12:55:48Z No. of bitstreams: 0<br>Made available in DSpace on 2016-12-21T12:55:48Z (GMT). No. of bitstreams: 0<br>O reator IEA-R1 do Instituto de Pesquisas Energéticas e Nucleares (IPENCNEN/ SP) é um reator de pesquisa tipo piscina aberta, projetado e construído pela empresa norte-americana \"Babcock & Wilcox\", tendo, como refrigerante e moderador, água leve deionizada e berílio e grafite como refletores. Até cerca de 1988, os sistemas de segurança do reator recebiam alimentação de uma única fonte de energia. Nos a

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OLIVEIRA, JOSE R. de. "Programa computacional para estudo da estrategia de controle de um reator nuclear do tipo PWR." reponame:Repositório Institucional do IPEN, 2002. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11060.

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11

Sun, Yaojun. "Heat transfer in molten core/concrete interaction systems." Diss., Georgia Institute of Technology, 1994. http://hdl.handle.net/1853/17105.

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12

RIBEIRO, MARIA A. M. "Estudo sobre sistemas de alivio da contencao aplicados a reatores de pequeno porte." reponame:Repositório Institucional do IPEN, 2001. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10959.

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13

Fehr, Brandon M. "Detailed study of the transient rod pneumatic system on the annular core research reactor." Thesis, Georgia Institute of Technology, 2016. http://hdl.handle.net/1853/55032.

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Throughout the history of the Annular Core Research Reactor (ACRR), Transient Rod (TR) A has experienced an increased rate of failure versus the other two TRs (B and C). Either by pneumatic force or electric motor, the transient rods remove the poison rods from the ACRR core allowing for the irradiation of experiments. In order to develop causes for why TR A is failing (rod break) more often, a better understanding of the whole TR system and its components is needed. This study aims to provide a foundational understanding of how the TR pneumatic system affects the motion of the TRs and the re

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14

HIROMOTO, MARIA Y. K. "PSINCO-um programa para calculo da distribuicao de potencia e supervisao do nucleo de reatores nucleares, utilizando sinais de detetores tipo 'SPD'." reponame:Repositório Institucional do IPEN, 1998. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10706.

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15

Delaney, Michael J. (Michael James) 1979. "Risk-informed design guidance for a Generation-IV gas-cooled fast reactor emergency core cooling system." Thesis, Massachusetts Institute of Technology, 2004. http://hdl.handle.net/1721.1/17747.

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Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2004.<br>Includes bibliographical references (p. 75-77).<br>Fundamental objectives of sustainability, economics, safety and reliability, and proliferation resistance, physical protection and stakeholder relations must be considered during the design of an advanced reactor. However, an advanced reactor's core damage frequency dominates all other considerations at the preliminary stage of reactor design. An iterative four-step methodology to guide the MIT gas-cooled fast reactor emergency core cooling system desi

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16

SOUZA, ROSE M. G. do P. "Determinação do fator de pico de potência utilizando barras de controle, detectores ex-core e redes neurais." reponame:Repositório Institucional do IPEN, 2005. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11296.

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17

Verwey, Aldo. "Modelling of a passive reactor cavity cooling system (RCCS) for a nuclear reactor core subject to environmental changes and the optimisation of the RCCS radiation heat shield heat shield." Thesis, Stellenbosch : University of Stellenbosch, 2010. http://hdl.handle.net/10019.1/4303.

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Thesis (MScEng (Mechanical and Mechatronic Engineering))--University of Stellenbosch, 2010.<br>ENGLISH ABSTRACT: A reactor cavity cooling system (RCCS) is used in the PBMR to protect the concrete citadel surrounding the reactor from direct nuclear radiation impingement and heat. The speci ed maximum operating temperature of the concrete structure is 65 ±C for normal operating conditions and 125 ±C for emergency shut-down conditions. A conceptual design of an entirely passive RCCS suitable for the PBMR was done by using closed loop thermosyphon heat pipes (CLTHPs) to remove heat from a rad

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18

STEFANI, GIOVANNI L. de. "Sobre a técnica de Rod Drop em medidas de reatividade integral em bancos de controle e segurança de reatores nucleares." reponame:Repositório Institucional do IPEN, 2013. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10210.

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GONCALVES, IRACI M. P. "Monitoração e diagnóstico para detecção de falhas de sensores utilizando a metodologia GMDH." reponame:Repositório Institucional do IPEN, 2006. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11382.

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20

Moloko, Lesego Ernest. "Impact of beryllium reflector ageing on Safari–1 reactor core parameters / L.E. Moloko." Thesis, North-West University, 2011. http://hdl.handle.net/10394/6930.

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The build–up of 6Li and 3He, that is, the strong thermal neutron absorbers or the so called "neutron poisons", in the beryllium reflector changes the physical characteristics of the reactor, such as reactivity, neutron spectra, neutron flux level, power distribution, etc.; furthermore,gaseous isotopes such as 3H and 4He induce swelling and embrittlement of the reflector. The SAFARI–1 research reactor, operated by Necsa at Pelindaba in South Africa, uses a beryllium reflector on three sides of the core, consisting of 19 beryllium reflector elements in total. This MTR went critical in 1965, and

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21

Xenofontos, Thalia. "Development of a dynamic stochastic neutronic code for the analysis of conventional and hybrid nuclear reactors." Thesis, Université Paris-Saclay (ComUE), 2018. http://www.theses.fr/2018SACLX013/document.

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La nécessité de simulations précises d’un réacteur nucléaire et spécialement dans des cas de cœurs et de configurations de combustible complexes, a imposé un usage accru de Codes Neutroniques Stochastiques (CNS). De plus, une demande a émergé pour des CNS à capacité inhérente d’estimation en continu de la variation de la composition isotopique du cœur ainsi qu’à couplage thermo-hydraulique optimisé. Des capacités supplémentaires sont exigées pour ces codes au vu de leur utilisation pour l’étude de nouveaux concepts de réacteur comme les Réacteurs Conduits par Accélérateur (RCA). Plus préciséme

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22

Ghasabyan, Levon. "Use of Serpent Monte-Carlo code for development of 3D full-core models of Gen-IV fast-spectrum reactors and preparation of group constants for transiet analyses with PARCS/TRACE coupled system." Thesis, KTH, Fysik, 2013. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-118072.

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23

Monti, Lanfranco [Verfasser]. "Multi-scale, coupled reactor physics - thermal-hydraulics system and applications to the HPLWR 3 pass core / von Lanfranco Monti." Eggenstein-Leopoldshafen : Forschungszentrum Karlsruhe GmbH, 2009. http://d-nb.info/100505844X/34.

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24

Singo, Thifhelimbilu Daphney. "Development of a high flux neutron radiation detection system for in-core temperature monitoring." Thesis, Stellenbosch : Stellenbosch University, 2012. http://hdl.handle.net/10019.1/19999.

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Thesis (PhD)--Stellenbosch University, 2012.<br>ENGLISH ABSTRACT: The objective of this research was to develop a neutron detection system that incorporates a mass spectrometer to measure high neutron flux in a nuclear reactor environment. This system consists of slow and fast neutron detector elements for measuring fluxes in those energy regions respectively. The detector should further be capable of withstanding the harsh conditions associated with a high temperature reactor. This novel detector which was initially intended for use in the PBMR reactor has possible applications as an i

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25

Zhao, Chengyuan. "Effect of Temperature on Biogas Production in Anaerobic Treatment of Domestic Wastewater UASB System in Hammarby Sjöstadsverk." Thesis, KTH, VA-teknik, Vatten, Avlopp och Avfall, 2011. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-95415.

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The upflow anaerobic sludge blanket (UASB) reactor is widely used around the world to treat variety of domestic andindustrial wastewater with three main advantages: production of biogas as renewable energy, no need of support structure for development of microorganisms and high rate treatment efficiency with low rate of biomass production. This study evaluates the effect of temperature on biogas production and CODtotal removal in Line 4-UASB system treating domestic wastewater in Hammarby Sjöstadsverk. Eight parameters were examined including the following: Influent and effluent temperature, p

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26

BAPTISTA, FILHO BENEDITO D. "Redes neurais para controle de sistemas de reatores nucleares." reponame:Repositório Institucional do IPEN, 1998. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10723.

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27

Gonçalves, Iraci Martinez Pereira. "Monitoração e diagnóstico para detecção de falhas de sensores utilizando a metodologia GMDH." Universidade de São Paulo, 2006. http://www.teses.usp.br/teses/disponiveis/85/85133/tde-04062012-144516/.

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O sistema de detecção de falhas e diagnóstico é um sistema de suporte ao operador dedicado a funções específicas que alertam os operadores para problemas de falhas em sensores e atuadores, e auxiliam no diagnóstico antes que os limites normais de alarmes sejam atingidos. Sistemas de suporte ao operador surgiram para diminuir a complexidade dos painéis causada pelo grande aumento de informação disponível nas salas de controle das centrais nucleares. Neste trabalho foi desenvolvido um Sistema de Monitoração e Diagnóstico utilizando a metodologia GMDH (Group Method of Data Handling) aplicado ao r

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28

Nalepa, Lauren Elizabeth. "Quantitative assessment of the impact of the previous cycle's core exposure on the transient response of boiling water reactor systems." [Gainesville, Fla.] : University of Florida, 2008. http://purl.fcla.edu/fcla/etd/UFE0022859.

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CARNEIRO, ALVARO L. G. "Desenvolvimento de sistema de monitoracao e diagnostico aplicado a valvulas moto-operadas utilizadas em centrais nucleares." reponame:Repositório Institucional do IPEN, 2003. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11109.

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SESINI, PAULA A. "Analise das instabilidades termo-hidraulicas em um circuito operando em regime de circulacao natural bi-fasico." reponame:Repositório Institucional do IPEN, 1998. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10686.

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Skwarcan-Bidakowski, Alexander. "Nuclear reactor core model for the advancednuclear fuel cycle simulator FANCSEE. Advanceduse of Monte Carlo methods in nuclear reactorcalculations." Thesis, Institutionen för Reaktorfysik, 2017. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-324260.

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A detailed reactor core modeling of the LOVIISA-2 PWR and FORSMARK-3BWR was performed in the Serpent 2 Continuous Energy Monte-Carlocode.Both models of the reactors were completed but the approximations ofthe atomic densities of nuclides present in the core differedsignificantly.In the LOVIISA-2 PWR, the predicted atomic density for the nuclidesapproximated by Chebyshev Rational Approximation method (CRAM)coincided with the corrected atomic density simulated by the Serpent2 program. In the case of FORSMARK-3 BWR, the atomic density fromCRAM poorly approximated the data returned by the simulati

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Reiter, Christian [Verfasser], Winfried [Akademischer Betreuer] Petry, Winfried [Gutachter] Petry, and Rafael [Gutachter] Macián-Juan. "A model system for transient calculations for research reactors with a compact core / Christian Reiter ; Gutachter: Winfried Petry, Rafael Macián-Juan ; Betreuer: Winfried Petry." München : Universitätsbibliothek der TU München, 2019. http://d-nb.info/1230552685/34.

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Negm, Hani Hussein. "Studies on the Optimum Geometry for a Nuclear Resonance Fluorescence Detection System for Nuclear Security Applications." Kyoto University, 2014. http://hdl.handle.net/2433/193589.

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Weiß, Frank-Peter. "Aktuelle Themen der Reaktorsicherheitsforschung in Deutschland." Forschungszentrum Dresden, 2010. http://nbn-resolving.de/urn:nbn:de:bsz:d120-qucosa-28406.

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Die Veranstaltung widmete sich mit der Borverdünnung in Druckwasserreaktoren bzw. mit der Verstopfung der Sumpfansaugsiebe durch freigesetztes Isolationsmaterial schwerpunktmäßig zwei Themen der Reaktorsicherheit, die auch in aktuellen Aufsichtsverfahren eine Rolle spielen. Eingebettet in den internationalen Kontext wollten die Veranstalter die sicherheitstechnische Bedeutung dieser Themen für die deutschen Anlagen beleuchten und die Auswirkungen auf die zu erbringenden Sicherheitsnachweise und den Anlagenbetrieb darstellen. Dabei kamen Gutachter, Vertreter der Forschung, Hersteller und Betrei

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Weiß, Frank-Peter. "Aktuelle Themen der Reaktorsicherheitsforschung in Deutschland." Forschungszentrum Rossendorf, 2006. https://hzdr.qucosa.de/id/qucosa%3A21667.

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Die Veranstaltung widmete sich mit der Borverdünnung in Druckwasserreaktoren bzw. mit der Verstopfung der Sumpfansaugsiebe durch freigesetztes Isolationsmaterial schwerpunktmäßig zwei Themen der Reaktorsicherheit, die auch in aktuellen Aufsichtsverfahren eine Rolle spielen. Eingebettet in den internationalen Kontext wollten die Veranstalter die sicherheitstechnische Bedeutung dieser Themen für die deutschen Anlagen beleuchten und die Auswirkungen auf die zu erbringenden Sicherheitsnachweise und den Anlagenbetrieb darstellen. Dabei kamen Gutachter, Vertreter der Forschung, Hersteller und Betrei

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Cekl, Jakub. "Model palivového souboru tlakovodního reaktoru západní koncepce." Master's thesis, Vysoké učení technické v Brně. Fakulta elektrotechniky a komunikačních technologií, 2018. http://www.nusl.cz/ntk/nusl-376896.

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Leopard, Mitchell L. "Loose Canon on Deck: How Contemporary Christians React to Media Portrayals of Faith, Beliefs, and Rituals." unrestricted, 2007. http://etd.gsu.edu/theses/available/etd-04302007-181623/.

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Thesis (M.A.)--Georgia State University, 2007.<br>Title from file title page. Timothy Renick, committee chair; Christopher White, Kathryn McClymond, committee members. Electronic text (106 p.) : digital, PDF file. Description based on contents viewed Dec. 31, 2007. Includes bibliographical references (p. 96-100). Filmography: p. 94-95.

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MAURIZI, ANNE. "Reactivite chimique a haute temperature dans le systeme (u, zr, fe, o). Contribution a l'etude de la zircone comme recuperateur de corium." Paris 6, 1996. http://www.theses.fr/1996PA066617.

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Notre travail s'est inscrit dans le cadre d'un programme de gestion des accidents nucleaires envisageant de placer sous la cuve des reacteurs un recuperateur destine a confiner le corium resultant de la fusion du coeur. D'apres l'etude bibliographique que nous avons realisee, la zircone stabilisee semble le materiau refractaire le mieux adapte pour remplir ce role, compte tenu des contraintes physico-chimiques, mecaniques et thermiques imposees au recuperateur. Nos recherches ont permis d'etablir la nature des interactions zircone/fer a haute temperature, et de determiner certaines donnees exp

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Sunnevik, Klas. "Comparison of MAAP and MELCOR : and evaluation of MELCOR as a deterministic tool within RASTEP." Thesis, Uppsala universitet, Tillämpad kärnfysik, 2014. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-233768.

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This master's thesis is an investigation and evaluation of MELCOR (a software tool for severe accident analyses regarding nuclear power plants), or more correctly of the (ASEA-Atom BWR 75) reactor model developed for version 1.8.6 of MELCOR. The main objective was to determine if MELCOR, with the reactor model in question, is able to produce satisfactory results in severe accident analyses compared to results made by MAAP, which is currently the only official software tool for this application in Sweden. The thesis work is related to the RASTEP project. This project has been carried out in sev

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Pavlíček, Michal. "Přístupy k zajištění jaderné bezpečnosti u reaktorů 3. generace." Master's thesis, Vysoké učení technické v Brně. Fakulta strojního inženýrství, 2010. http://www.nusl.cz/ntk/nusl-229286.

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The main target of the master´s thesis is reviewing the generation III nuclear reactors in term of the nuclear safety. At first we have to learn some theory of the nuclear safety in order to understand safety systems of the generation III nuclear reactors. Therefore the thesis is divided into two parts. Legislative and technical approaches to nuclear safety are mentioned in the first part. Regulatory bodies, whose task is to supervise nuclear safety in the nuclear power plants, belongs to the legislative approaches. There are defined terms such as defence in depth, redundancy, diversity, etc.

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Kuo, Chiao-Chun, and 郭巧君. "Assessment of RELAP5/MOD3 System Thermal Hydraulic Analysis Code Using Power Test Data of A Westinghouse PWR Reactor." Thesis, 1997. http://ndltd.ncl.edu.tw/handle/85725665647038553204.

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Abdul-hamid, Shahab A. "Monte Carlo burnup analysis code development and application to an incore thermionic space nuclear power system." Thesis, 1993. http://hdl.handle.net/1957/35833.

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Lattice bum-up calculations in thermal reactors are complicated by the necessity for use of transport theory to represent fuel rods, control rods, and burnable absorbers, by many time-dependent variables which must be considered in the analysis, and by geometric complexity which introduces time-dependent, spatial-spectral variations. Representation of lattice structure in a core is further complicated by fuel materials and loading patterns which can be non-symmetric, and by the type of material used as the moderator. The incore thermionic reactor system developed under the Advanced Thermionic

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(7473860), Dongli Huang. "Efficient Uncertainty Characterization Framework in Neutronics Core Simulation with Application to Thermal-Spectrum Reactor Systems." Thesis, 2020.

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<div>This dissertation is devoted to developing a first-of-a-kind uncertainty characterization framework (UCF) providing comprehensive, efficient and scientifically defendable methodologies for uncertainty characterization (UC) in best-estimate (BE) reactor physics simulations. The UCF is designed with primary application to CANDU neutronics calculations, but could also be applied to other thermal-spectrum reactor systems. The overarching goal of the UCF is to propagate and prioritize all sources of uncertainties, including those originating from nuclear data uncertainties, modeling assumption

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Colpo, Sarah E. "Pressurizer surge line Counter Current Flow Limitation during AP600 Mode 5 Cold Shutdown." Thesis, 1999. http://hdl.handle.net/1957/33578.

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Counter Current Flow Limitation (CCFL) was observed in the pressurizer surge line of the Oregon State University APEX facility during test NRC-10. This test simulated a one-inch diameter cold leg break with a failure of three of four of the fourth-stage Automatic Depressurization System (ADS) valves. The result was a high vapor flow rate through ADS 1-3, that caused CCFL in the pressurizer surge line and liquid holdup in the pressurizer. Because this liquid was not available for core cooling, further study of the passive safety systems in the AP600 under Mode 5 Cold Shutdown conditions was dee

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