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This research investigates the performance of new radiografted materials as adsorbents for ionic and colloidal 110 mAg, a prominent pollutant in pressurized water reactor (PWR) wastewater. A comparative analysis was conducted to evaluate their effectiveness against commercially available resins. Due to the increasing demand for new nuclear power plants and the growing global dependence on nuclear energy, it is crucial to minimize radioactive waste production. The study was conducted on simulated inactive and radioactive waste samples with trace amounts of 110 mAg that are hard or impossible to detect by standard analytic techniques. The synthesized materials were labeled PGE (amine functional groups) and PGM (sulfhydryl and amine functional groups). Functionalization with amine and sulfhydryl groups was chosen since it is well known that these groups strongly interact with Ag in both ionic and colloidal form. The tested materials demonstrate exceptional performance as adsorbents by completely removing (100 %) both ionic (PGM) and colloidal (PGE) forms of 110 mAg. Additionally, depending on the adsorbed species excellent elution of 110 mAg was achieved (>94 % for ions and >50 % for colloids) implying possible material reusability. Manufacturing cost of such adsorbents is orders of magnitude lower per kg (10 – 50 $/kg), compared to prices of commercial resins (100 – 10000 $/kg). These results indicate that such inexpensive materials could potentially be used in an efficient and selective way to treat PWR wastewater for 110 mAg contamination while reducing the operating cost and minimizing resulting solid waste.

A Loss-of-Coolant Accident (LOCA) represents a critical scenario for the safety of a nuclear reactor, as the loss of the reactor coolant can result in overheating and irreversible damages to the core. For the licensing of a nuclear reactor or fuel to be approved, multiple safety criterions are required to be fulfilled. One criteria is that the peak cladding temperatures for fuel rods can’t exceed the upper limit of 1204.4 degrees Celsius to prevent any damage on the fuel and dispersal of radioactive particles. In collaboration with Westinghouse Electric Sweden AB, the aim of this report is to update and improve Westinghouse’s thermal hydraulic codes used for calculating and conducting safety analysis of a reactor core environment during a hypothetical LOCA. The Westinghouse LOCA experiment G-2 were simulated in the thermal hydraulic codes GOBLIN and CHACHA, to evaluate if the codes can produce results that are similar to the G-2 results. Accurate models and methodologies for post-LOCA environments are necessary for the licensing process of the newly developed TRITON11-fuel. With GOBLIN and CHAHA the G-2 test were simulated and the surface temperatures of the fuel rods between the codes and the G-2 data were compared. CHACHA manages to simulate the initial heat-up process well, with 83 % of the simulated runs having an accepted surface temperature at 60 seconds into the test sequence. However, the cooling transient phase were difficult for CHACHA to provide results that were similar to G-2, as only one of the six simulated runs managed to fulfill the set requirements of the surface temperatures. The setup parameters for the six simulated runs were observed to have a large impact on the results, as the runs with the largest upper plenum pressure provided the best results. Several points of possible causes to the results were identified for further improvements of the thermal hydraulic codes, namely the calculation of heat transfer coefficients, the cooling calculation

A Loss-of-Coolant Accident (LOCA) represents a critical scenario for the safety of a nuclear reactor, as the loss of the reactor coolant can result in overheating and irreversible damages to the core. For the licensing of a nuclear reactor or fuel to be approved, multiple safety criterions are required to be fulfilled. One criteria is that the peak cladding temperatures for fuel rods can’t exceed the upper limit of 1204.4 degrees Celsius to prevent any damage on the fuel and dispersal of radioactive particles. In collaboration with Westinghouse Electric Sweden AB, the aim of this report is to update and improve Westinghouse’s thermal hydraulic codes used for calculating and conducting safety analysis of a reactor core environment during a hypothetical LOCA. The Westinghouse LOCA experiment G-2 were simulated in the thermal hydraulic codes GOBLIN and CHACHA, to evaluate if the codes can produce results that are similar to the G-2 results. Accurate models and methodologies for post-LOCA environments are necessary for the licensing process of the newly developed TRITON11-fuel. With GOBLIN and CHAHA the G-2 test were simulated and the surface temperatures of the fuel rods between the codes and the G-2 data were compared. CHACHA manages to simulate the initial heat-up process well, with 83 % of the simulated runs having an accepted surface temperature at 60 seconds into the test sequence. However, the cooling transient phase were difficult for CHACHA to provide results that were similar to G-2, as only one of the six simulated runs managed to fulfill the set requirements of the surface temperatures. The setup parameters for the six simulated runs were observed to have a large impact on the results, as the runs with the largest upper plenum pressure provided the best results. Several points of possible causes to the results were identified for further improvements of the thermal hydraulic codes, namely the calculation of heat transfer coefficients, the cooling calculation

There are currently two operating pressurized water reactors in Sweden, currently planning for 60 years of operation until 2041 and 2043. The acceptance of operation time is continuously evaluated at least every 10 years in a comprehensive mandatory periodic safety review that requires the utilities to continuously update and implement the developments in science and technology. The RPV welds have been shown by the applied surveillance program to be the limiting material for operation. The welds are manufactured according to the same specifications with a chemical composition with high nickel and manganese content. The welds show a large increase in transition temperature shift with an almost linear relationship to neutron fluence that is underestimated by most of the established embrittlement trend curves (ETCs). The current regulations from the Swedish Radiation Safety Authority are in general not detailed and prescriptive and hence permit plant-specific ETCs if they are sufficiently justified and based on proper material and plant conditions. This paper describes the bases for the ETCs and an ongoing work to revise the ETCs to enable the use of a material-specific master curve for crack initiation, KIC, with compliance with the reactor vessel integrity analyses., Part of ISBN 9780803177413QC 20230831

This thesis studied the performance of the thermal hydraulic software TRACE applied on an intermediate sized break (IB) happening on the cold leg in a pressurized water reactor (PWR), causing a loss-of-coolant accident (LOCA). The same accident has previously been simulated in the PKL Test Facility, which is a scaled version of a PWR and is used to simulate transients stemming from different accidents. The thesis was performed as a semi-blind study: firstly, the accident was simulated without any knowledge of the PKL results. When a final blind model was chosen, the PKL results were revealed, and the TRACE model was improved. Before the simulations of the IB-LOCA took place, the new internal parts in the upper parts of the reactor pressure vessel in PKL had to be modelled, and the steady state had to be tuned to attain the correct initial conditions. The simulations were performed by using the software SNAP together with TRACE, providing a graphical interface. TRACE achieved steady state with satisfying results regarding water levels, pressure losses and mass flows. The temperatures in TRACE deviated from the PKL temperatures but an explanation is uncertainties in PKL. To verify TRACE’s core output power, the calculation of the power was done by using mass flow rate and specific entropy and comparing to the heaters’ specified power. This resulted in lower output power meaning that the coolant was not heated enough. This indicated non-physical energy losses in the TRACE model and should be further investigated.The blind transient simulation, modelled with default choked flow and no offtake model, resulted in TRACE overestimating the break mass flow and the peak cladding temperatures, compared to the PKL reference solution. This resulted in the pressure decreasing too quickly and too early activation of the safety system. The modified simulations showed that it is important that the offtake model, which accounts for different flow regimes, is activated. Default choked

This thesis studied the performance of the thermal hydraulic software TRACE applied on an intermediate sized break (IB) happening on the cold leg in a pressurized water reactor (PWR), causing a loss-of-coolant accident (LOCA). The same accident has previously been simulated in the PKL Test Facility, which is a scaled version of a PWR and is used to simulate transients stemming from different accidents. The thesis was performed as a semi-blind study: firstly, the accident was simulated without any knowledge of the PKL results. When a final blind model was chosen, the PKL results were revealed, and the TRACE model was improved. Before the simulations of the IB-LOCA took place, the new internal parts in the upper parts of the reactor pressure vessel in PKL had to be modelled, and the steady state had to be tuned to attain the correct initial conditions. The simulations were performed by using the software SNAP together with TRACE, providing a graphical interface. TRACE achieved steady state with satisfying results regarding water levels, pressure losses and mass flows. The temperatures in TRACE deviated from the PKL temperatures but an explanation is uncertainties in PKL. To verify TRACE’s core output power, the calculation of the power was done by using mass flow rate and specific entropy and comparing to the heaters’ specified power. This resulted in lower output power meaning that the coolant was not heated enough. This indicated non-physical energy losses in the TRACE model and should be further investigated.The blind transient simulation, modelled with default choked flow and no offtake model, resulted in TRACE overestimating the break mass flow and the peak cladding temperatures, compared to the PKL reference solution. This resulted in the pressure decreasing too quickly and too early activation of the safety system. The modified simulations showed that it is important that the offtake model, which accounts for different flow regimes, is activated. Default choked

In the second half of the 20th century, in a phase of openness and growth after years of repression and poverty under an authoritarian government, Spain chose nuclear energy as a source of electricity generation by importing several nuclear power reactor designs. One of the nuclear plants requested was the Lemóniz plant, in the Basque Country, with a pressurized water reactor (PWR) design. The construction of this nuclear power plant promoted a social and cultural movement against technology and in favor of protecting the environment in the country and the identity values of the Basque region. In addition, the intrusion of the terrorist group ETA into the controversy and the beginning of the transition to democracy reinforced the decision in favor of a moratorium for all nuclear projects proposed in Spain., Na segunda metade do século passado, num estágio de abertura e crescimento após anos de repressão e pobreza sob um governo autoritário, a Espanha optou pela energia nuclear como fonte de geração de energia elétrica, importando vários projetos de reatores nucleares. Uma das usinas nucleares solicitadas foi Lemóniz, no País Basco, com um projeto de reator de água pressurizada (PWR, por suas siglas em inglês). A construção desta usina defendeu um movimento contra essa tecnologia e a favor da proteção do meio ambiente no país e dos valores de identidade da região basca. Além disso, o surgimento do grupo terrorista ETA na controvérsia e o início do estágio de transição para a democracia reforçaram a tomada de decisões políticas em favor de uma moratória para todos os projetos nucleares propostos na Espanha., En la segunda mitad del pasado siglo, en una etapa de apertura y de crecimiento tras años de represión y pobreza bajo un gobierno autoritario, España apostó por la energía nuclear como fuente de generación de energía eléctrica, importando varios diseños de reactores nucleares. Una de las centrales nucleares solicitadas fue Lemóniz, en el País Vasco, con diseño de reactor de agua a presión (pressurized water reactor – PWR). La construcción de esta central abanderó un movimiento social y cultural en contra de esta tecnología y a favor de la protección del medioambiente en el país y de los valores identitarios de la región vasca. Además, la irrupción de la banda terrorista ETA en la controversia y el comienzo de la etapa de transición hacia la democracia reforzaron la toma de una decisión política a favor de una moratoria para todos los proyectos nucleares propuestos en España.

In the second half of the 20th century, in a phase of openness and growth after years of repression and poverty under an authoritarian government, Spain chose nuclear energy as a source of electricity generation by importing several nuclear power reactor designs. One of the nuclear plants requested was the Lemóniz plant, in the Basque Country, with a pressurized water reactor (PWR) design. The construction of this nuclear power plant promoted a social and cultural movement against technology and in favor of protecting the environment in the country and the identity values of the Basque region. In addition, the intrusion of the terrorist group ETA into the controversy and the beginning of the transition to democracy reinforced the decision in favor of a moratorium for all nuclear projects proposed in Spain., Na segunda metade do século passado, num estágio de abertura e crescimento após anos de repressão e pobreza sob um governo autoritário, a Espanha optou pela energia nuclear como fonte de geração de energia elétrica, importando vários projetos de reatores nucleares. Uma das usinas nucleares solicitadas foi Lemóniz, no País Basco, com um projeto de reator de água pressurizada (PWR, por suas siglas em inglês). A construção desta usina defendeu um movimento contra essa tecnologia e a favor da proteção do meio ambiente no país e dos valores de identidade da região basca. Além disso, o surgimento do grupo terrorista ETA na controvérsia e o início do estágio de transição para a democracia reforçaram a tomada de decisões políticas em favor de uma moratória para todos os projetos nucleares propostos na Espanha., En la segunda mitad del pasado siglo, en una etapa de apertura y de crecimiento tras años de represión y pobreza bajo un gobierno autoritario, España apostó por la energía nuclear como fuente de generación de energía eléctrica, importando varios diseños de reactores nucleares. Una de las centrales nucleares solicitadas fue Lemóniz, en el País Vasco, con diseño de reactor de agua a presión (pressurized water reactor – PWR). La construcción de esta central abanderó un movimiento social y cultural en contra de esta tecnología y a favor de la protección del medioambiente en el país y de los valores identitarios de la región vasca. Además, la irrupción de la banda terrorista ETA en la controversia y el comienzo de la etapa de transición hacia la democracia reforzaron la toma de una decisión política a favor de una moratoria para todos los proyectos nucleares propuestos en España.

This paper presents a sensitivity study of nodal scheme in MELCOR simulation of severe accidents in a pressurized water reactor, with the objective to estimate the nodal effects on some in-vessel and exvessel processes and phenomena, including thermal-hydraulic response, core degradation and relocation, hydrogen generation, vessel failure, containment pressurization and venting, source term. For this purpose, three nodal schemes (i.e., coarse, medium and fine meshes) of the COR package of the MELCOR code are chosen to analyze two severe accident scenarios: small break loss of coolant accident (SBLOCA) and large break loss-of-coolant accident (LBLOCA), both combined with station blackout. The results show that the nodal schemes mainly affect the calculations of heat transfers from the core to coolant and heat structures, relatively affecting the core degradation and relocation to the lower head of the reactor pressure vessel. As for the consequences, the coarse mesh tends to predict slower core relocation progressions and a later failure of RPV lower head. Moreover, more hydrogen generation by cladding oxidation can be observed in the coarse mesh case. The nodal schemes have little impact on the estimation of in-containment source term. Meanwhile, the simulations with fine mesh may also provide more detailed distributions of corium masses and temperatures, as well as heat fluxes, affecting thermal and mechanic behavior of RPV lower head., QC 20210710

In the second half of the 20th century, in a phase of openness and growth after years of repression and poverty under an authoritarian government, Spain chose nuclear energy as a source of electricity generation by importing several nuclear power reactor designs. One of the nuclear plants requested was the Lemóniz plant, in the Basque Country, with a pressurized water reactor (PWR) design. The construction of this nuclear power plant promoted a social and cultural movement against technology and in favor of protecting the environment in the country and the identity values of the Basque region. In addition, the intrusion of the terrorist group ETA into the controversy and the beginning of the transition to democracy reinforced the decision in favor of a moratorium for all nuclear projects proposed in Spain., Na segunda metade do século passado, num estágio de abertura e crescimento após anos de repressão e pobreza sob um governo autoritário, a Espanha optou pela energia nuclear como fonte de geração de energia elétrica, importando vários projetos de reatores nucleares. Uma das usinas nucleares solicitadas foi Lemóniz, no País Basco, com um projeto de reator de água pressurizada (PWR, por suas siglas em inglês). A construção desta usina defendeu um movimento contra essa tecnologia e a favor da proteção do meio ambiente no país e dos valores de identidade da região basca. Além disso, o surgimento do grupo terrorista ETA na controvérsia e o início do estágio de transição para a democracia reforçaram a tomada de decisões políticas em favor de uma moratória para todos os projetos nucleares propostos na Espanha., En la segunda mitad del pasado siglo, en una etapa de apertura y de crecimiento tras años de represión y pobreza bajo un gobierno autoritario, España apostó por la energía nuclear como fuente de generación de energía eléctrica, importando varios diseños de reactores nucleares. Una de las centrales nucleares solicitadas fue Lemóniz, en el País Vasco, con diseño de reactor de agua a presión (pressurized water reactor – PWR). La construcción de esta central abanderó un movimiento social y cultural en contra de esta tecnología y a favor de la protección del medioambiente en el país y de los valores identitarios de la región vasca. Además, la irrupción de la banda terrorista ETA en la controversia y el comienzo de la etapa de transición hacia la democracia reforzaron la toma de una decisión política a favor de una moratoria para todos los proyectos nucleares propuestos en España.

The German software tool ATHLET (Analysis of THermohydraulics of Leaks and Transients) is continuously being developed for the simulation of the nuclear power plant behaviour in the event of transients and accidents. The focus of a current joint research project named „ATHLET Modul Zinkborat“ (AZora) is the development and validation of an ATHLET module on the basis of the current state of research on chemical long-term effects according to PWR LOCA. The module is intended to simulate thermohydraulic effects of zinc borate precipitations in the reactor core originating from long-term corrosion processes in the reactor sump during sump recirculation operation after postulated loss-of-coolant accidents in PWR. To develop and validate the sub-models, generic experiments at lab-scale as well as at semi-technical scale are planned to be carried out in unique test facilities. The structure of the module as well as the experimental strategies and the related facilities are described in the article.

The experiment ROCOM E2.3 represents a boron dilution event in a KONVOI-type pressurized lightwater reactor. It was conducted at room temperature with de-mineralized water without boric acid. Underborated water slugs were modelled by adding Ethanol in order to adjust a density difference of 1.22% with respect to the regular coolant inventory. At the beginning of the experiment, the slugs are enclosed between two valves in the cold legs of loops 1 and 2. The volume of the two water slugs accounts for 0.0576 m 3 (57.6 l) each and the slug fronts are located at 1.8 m upstream of the pressure vessel inlet nozzles. The experiment is started by opening the loop valves and running up the circulation pumps. The time dependency of the volumetric flow rates in all four coolant loops can be found in https://doi.org/10.1016/j.nucengdes.2020.110776. During the experiment, the mixing process was recorded by wire-mesh conductivity sensors at various positions within the coolant loops and the pressure vessel.  The nomenclature of the data files as well as the format of the tables are described in the accompanying document DataDescription_ROCOME23.pdf.

[Abstract] Nuclear energy is presented as a real option in the face of the current problem of climate change and the need to reduce CO2 emissions. The nuclear reactor design with the greatest global impact throughout history and which has the most ambitious development plans is the Pressurized Water Reactor (PWR). Thus, a global review of such a reactor design is presented in this paper, utilizing the analysis of (i) technical aspects of the different variants of the PWR design implemented over the past eight years, (ii) the level of implementation of PWR nuclear power plants in the world, and (iii) a life extension scenario and future trends in PWR design based on current research and development (R&D) activity. To develop the second analysis, a statistical study of the implementation of the different PWR variants has been carried out. Such a statistical analysis is based on the operating factor, which represents the relative frequency of reactors operating around the world. The results reflect the hegemony of the western variants in the 300 reactors currently operating, highlighting the North American and French versions. Furthermore, a simulation of a possible scenario of increasing the useful life of operational PWRs up to 60 years has been proposed, seeing that in 2050 the generation capacity of nuclear PWRs power plants will decrease by 50%, and the number of operating reactors by 70%.

In a joint research project of the Zittau/Goerlitz University of Applied Sciences (HSZG), the Technische Universitaet Dresden (TUD) and the Helmholtz-Zentrum Dresden-Rossendorf (HZDR), the main emphasis is the time-related assignment of simultaneous and interacting mechanisms at zinc sources and zinc sinks at boundary conditions of a loss-of-coolant accident (LOCA) in German pressurized water reactors (PWR). The according experiments are carried out at semi-technical as well as at laboratory scale. Zinc is used as a protective coating, e.g. for gratings in the containment, showing high corrosion resistance due to a gradual formation of passivating layers. In contrast, its long-term behaviour during LOCA changes significantly under the influence of the coolant chemistry of German PWR. As a consequence, according installations in the containment act as zinc sources due to corrosion. Released zinc ions change the chemical properties of the coolant and could e.g. lead to layer-forming depositions of zinc borates in the core, which increase the possibility of a hindered heat dissipation. For experimental and methodical investigations of these phenomena, the test rig “Zittau Flow Tray” (ZFT), a scaled sump model of a German PWR, was equipped with a full-length 3×3 fuel assembly (FA) dummy acting as core model, a preheater and a cooler component. Nine 4.4 m long fuel rod dummies simulate the decay heat by internal heating cartridges. This rig design enables experimental investigation of physico-chemical mechanisms considering coolant containing boric acid and zinc and their influence on the thermo-hydraulic processes in the reactor core at post-LOCA boundary conditions. The time depending zinc release at hot-dip galvanized gratings (HGG) was investigated regarding their position (e.g. inside or near the leaking jet, freely suspended or submerged in the coolant) and their surface area as well as temperature and flow rate of the coolant. The experimental database allows the

In a joint research project of the Zittau/Goerlitz University of Applied Sciences (HSZG), the Technische Universitaet Dresden (TUD) and the Helmholtz-Zentrum Dresden-Rossendorf (HZDR), the main emphasis is the time-related assignment of simultaneous and interacting mechanisms at zinc sources and zinc sinks at boundary conditions of a loss-of-coolant accident (LOCA) in German pressurized water reactors (PWR). The according experiments are carried out at semi-technical as well as at laboratory scale. Zinc is used as a protective coating, e.g. for gratings in the containment, showing high corrosion resistance due to a gradual formation of passivating layers. In contrast, its long-term behaviour during LOCA changes significantly under the influence of the coolant chemistry of German PWR. As a consequence, according installations in the containment act as zinc sources due to corrosion. Released zinc ions change the chemical properties of the coolant and could e.g. lead to layer-forming depositions of zinc borates in the core, which increase the possibility of a hindered heat dissipation. For experimental and methodical investigations of these phenomena, the test rig “Zittau Flow Tray” (ZFT), a scaled sump model of a German PWR, was equipped with a full-length 3×3 fuel assembly (FA) dummy acting as core model, a preheater and a cooler component. Nine 4.4 m long fuel rod dummies simulate the decay heat by internal heating cartridges. This rig design enables experimental investigation of physico-chemical mechanisms considering coolant containing boric acid and zinc and their influence on the thermo-hydraulic processes in the reactor core at post-LOCA boundary conditions. The time depending zinc release at hot-dip galvanized gratings (HGG) was investigated regarding their position (e.g. inside or near the leaking jet, freely suspended or submerged in the coolant) and their surface area as well as temperature and flow rate of the coolant. The experimental database allows the

In a joint research project of the Zittau/Goerlitz University of Applied Sciences (HSZG), the Technische Universitaet Dresden (TUD) and the Helmholtz-Zentrum Dresden-Rossendorf (HZDR), the main emphasis is the time-related assignment of simultaneous and interacting mechanisms at zinc sources and zinc sinks at boundary conditions of a loss-of-coolant accident (LOCA) in German pressurized water reactors (PWR). The according experiments are carried out at semi-technical as well as at laboratory scale. Zinc is used as a protective coating, e.g. for gratings in the containment, showing high corrosion resistance due to a gradual formation of passivating layers. In contrast, its long-term behaviour during LOCA changes significantly under the influence of the coolant chemistry of German PWR. As a consequence, according installations in the containment act as zinc sources due to corrosion. Released zinc ions change the chemical properties of the coolant and could e.g. lead to layer-forming depositions of zinc borates in the core, which increase the possibility of a hindered heat dissipation. For experimental and methodical investigations of these phenomena, the test rig “Zittau Flow Tray” (ZFT), a scaled sump model of a German PWR, was equipped with a full-length 3×3 fuel assembly (FA) dummy acting as core model, a preheater and a cooler component. Nine 4.4 m long fuel rod dummies simulate the decay heat by internal heating cartridges. This rig design enables experimental investigation of physico-chemical mechanisms considering coolant containing boric acid and zinc and their influence on the thermo-hydraulic processes in the reactor core at post-LOCA boundary conditions. The time depending zinc release at hot-dip galvanized gratings (HGG) was investigated regarding their position (e.g. inside or near the leaking jet, freely suspended or submerged in the coolant) and their surface area as well as temperature and flow rate of the coolant. The experimental database allows the

Zur Abfuhr der Nachzerfallswärme in der Spätphase eines Kühlmittelverluststörfalles in Druckwasserreaktoren wird das aus den Leck im Primärkreislauf austretende Kühlwasser aus dem Reaktorsumpf im sog. Sumpfumwälzbetrieb mittels der Niederdruckeinspeisepumpen in den Reaktorkern rezirkuliert. Im Containment kommt das Kühlmittel dabei in Kontakt mit Fremdstoffen, wie z.B. Isoliermaterialfasern, Staub und korrosiven Materialien, welche einerseits die Kühlmittelchemie und andererseits die Performance der den Pumpen vorgeschalteten Sumpfsiebe beeinflussen können. Weiterhin haben Studien gezeigt, dass feuerverzinkte Containment-Einbauten (z.B. Lichtgitterroste, Stützgitter von Sumpfsieben, Kanäle) einer beschleunigten Korrosion durch das borsäurehaltige Kühlmittel unterliegen. Die daraus resultierenden thermohydraulischen Effekte hängen in hohem Maße vom Löslichkeitsverhalten der Korrosionsprodukte ab. So können unlösliche Korrosionspartikel zu einem erhöhten Differenzdruck an den bereits mit Isoliermaterialfasern beladenen Sumpfsieben führen, während lösliche Korrosionsprodukte nicht zurückgehalten werden und somit in den Kern gelangen, was unter Umständen in Ausfällungsprozessen durch Temperaturänderungen resultiert. Da ein Einfluss dieser Effekte auf die Kernkühlung nicht ausgeschlossen werden kann, ist die Untersuchung der zugrundeliegenden physikochemischen Korrosions-, Ausfällungs- und Ablagerungsprozesse sowie deren thermohydraulischen Folgen Gegenstand von gemeinsamen Forschungsvorhaben des Helmholtz-Zentrums Dresden-Rossendorf, der TU Dresden sowie der Hochschule Zittau-Görlitz. Der Vortrag gibt einen Überblick über die bisherigen Forschungsarbeiten der o.g. Institutionen sowie die wesentlichen Ergebnisse der entsprechenden BMWi-Forschungsvorhaben im Kontext der Reaktorsicherheitsforschung.

In a joint research project of the Zittau/Goerlitz University of Applied Sciences (HSZG), the Technische Universitaet Dresden (TUD) and the Helmholtz-Zentrum Dresden-Rossendorf (HZDR), the main emphasis is the time-related assignment of simultaneous and interacting mechanisms at zinc sources and zinc sinks at boundary conditions of a loss-of-coolant accident (LOCA) in German pressurized water reactors (PWR). The according experiments are carried out at semi-technical as well as at laboratory scale. Zinc is used as a protective coating, e.g. for gratings in the containment, showing high corrosion resistance due to a gradual formation of passivating layers. In contrast, its long-term behaviour during LOCA changes significantly under the influence of the coolant chemistry of German PWR. As a consequence, according installations in the containment act as zinc sources due to corrosion. Released zinc ions change the chemical properties of the coolant and could e.g. lead to layer-forming depositions of zinc borates in the core, which increase the possibility of a hindered heat dissipation. For experimental and methodical investigations of these phenomena, the test rig “Zittau Flow Tray” (ZFT), a scaled sump model of a German PWR, was equipped with a full-length 3×3 fuel assembly (FA) dummy acting as core model, a preheater and a cooler component. Nine 4.4 m long fuel rod dummies simulate the decay heat by internal heating cartridges. This rig design enables experimental investigation of physico-chemical mechanisms considering coolant containing boric acid and zinc and their influence on the thermo-hydraulic processes in the reactor core at post-LOCA boundary conditions. The time depending zinc release at hot-dip galvanized gratings (HGG) was investigated regarding their position (e.g. inside or near the leaking jet, freely suspended or submerged in the coolant) and their surface area as well as temperature and flow rate of the coolant. The experimental database allows the

In a joint research project of the Zittau/Goerlitz University of Applied Sciences (HSZG), the Technische Universitaet Dresden (TUD) and the Helmholtz-Zentrum Dresden-Rossendorf (HZDR), the main emphasis is the time-related assignment of simultaneous and interacting mechanisms at zinc sources and zinc sinks at boundary conditions of a loss-of-coolant accident (LOCA) in German pressurized water reactors (PWR). The according experiments are carried out at semi-technical as well as at laboratory scale. Zinc is used as a protective coating, e.g. for gratings in the containment, showing high corrosion resistance due to a gradual formation of passivating layers. In contrast, its long-term behaviour during LOCA changes significantly under the influence of the coolant chemistry of German PWR. As a consequence, according installations in the containment act as zinc sources due to corrosion. Released zinc ions change the chemical properties of the coolant and could e.g. lead to layer-forming depositions of zinc borates in the core, which increase the possibility of a hindered heat dissipation. For experimental and methodical investigations of these phenomena, the test rig “Zittau Flow Tray” (ZFT), a scaled sump model of a German PWR, was equipped with a full-length 3×3 fuel assembly (FA) dummy acting as core model, a preheater and a cooler component. Nine 4.4 m long fuel rod dummies simulate the decay heat by internal heating cartridges. This rig design enables experimental investigation of physico-chemical mechanisms considering coolant containing boric acid and zinc and their influence on the thermo-hydraulic processes in the reactor core at post-LOCA boundary conditions. The time depending zinc release at hot-dip galvanized gratings (HGG) was investigated regarding their position (e.g. inside or near the leaking jet, freely suspended or submerged in the coolant) and their surface area as well as temperature and flow rate of the coolant. The experimental database allows the

A statistical approach is used to analyse the early phase of station blackout accident for generic German PWR with the best estimate system code ATHLET-CD as a computation tool. The analysis is mainly focused on the timescale uncertainties of the accident events which can be detected at the plant. The developed input deck allows variations of all input uncertainty parameters relevant to the case. The list of identified and quantified input uncertainties includes 30 parameters related to the simulated physical phenomena/processes. Time uncertainties of main events as well as the major contributors to these uncertainties are defined. A linear regression analysis is used for predicting times of future events from detected times of occurred/past events. The presented statistical approach could be helpful for assessing and improving existing or elaborating additional emergency operating procedures aimed to prevent severe damage of reactor core.

Zur Abfuhr der Nachzerfallswärme in der Spätphase eines Kühlmittelverluststörfalles in Druckwasserreaktoren wird das aus den Leck im Primärkreislauf austretende Kühlwasser aus dem Reaktorsumpf im sog. Sumpfumwälzbetrieb mittels der Niederdruckeinspeisepumpen in den Reaktorkern rezirkuliert. Im Containment kommt das Kühlmittel dabei in Kontakt mit Fremdstoffen, wie z.B. Isoliermaterialfasern, Staub und korrosiven Materialien, welche einerseits die Kühlmittelchemie und andererseits die Performance der den Pumpen vorgeschalteten Sumpfsiebe beeinflussen können. Weiterhin haben Studien gezeigt, dass feuerverzinkte Containment-Einbauten (z.B. Lichtgitterroste, Stützgitter von Sumpfsieben, Kanäle) einer beschleunigten Korrosion durch das borsäurehaltige Kühlmittel unterliegen. Die daraus resultierenden thermohydraulischen Effekte hängen in hohem Maße vom Löslichkeitsverhalten der Korrosionsprodukte ab. So können unlösliche Korrosionspartikel zu einem erhöhten Differenzdruck an den bereits mit Isoliermaterialfasern beladenen Sumpfsieben führen, während lösliche Korrosionsprodukte nicht zurückgehalten werden und somit in den Kern gelangen, was unter Umständen in Ausfällungsprozessen durch Temperaturänderungen resultiert. Da ein Einfluss dieser Effekte auf die Kernkühlung nicht ausgeschlossen werden kann, ist die Untersuchung der zugrundeliegenden physikochemischen Korrosions-, Ausfällungs- und Ablagerungsprozesse sowie deren thermohydraulischen Folgen Gegenstand von gemeinsamen Forschungsvorhaben des Helmholtz-Zentrums Dresden-Rossendorf, der TU Dresden sowie der Hochschule Zittau-Görlitz. Der Vortrag gibt einen Überblick über die bisherigen Forschungsarbeiten der o.g. Institutionen sowie die wesentlichen Ergebnisse der entsprechenden BMWi-Forschungsvorhaben im Kontext der Reaktorsicherheitsforschung.

Zur Abfuhr der Nachzerfallswärme in der Spätphase eines Kühlmittelverluststörfalles in Druckwasserreaktoren wird das aus den Leck im Primärkreislauf austretende Kühlwasser aus dem Reaktorsumpf im sog. Sumpfumwälzbetrieb mittels der Niederdruckeinspeisepumpen in den Reaktorkern rezirkuliert. Im Containment kommt das Kühlmittel dabei in Kontakt mit Fremdstoffen, wie z.B. Isoliermaterialfasern, Staub und korrosiven Materialien, welche einerseits die Kühlmittelchemie und andererseits die Performance der den Pumpen vorgeschalteten Sumpfsiebe beeinflussen können. Weiterhin haben Studien gezeigt, dass feuerverzinkte Containment-Einbauten (z.B. Lichtgitterroste, Stützgitter von Sumpfsieben, Kanäle) einer beschleunigten Korrosion durch das borsäurehaltige Kühlmittel unterliegen. Die daraus resultierenden thermohydraulischen Effekte hängen in hohem Maße vom Löslichkeitsverhalten der Korrosionsprodukte ab. So können unlösliche Korrosionspartikel zu einem erhöhten Differenzdruck an den bereits mit Isoliermaterialfasern beladenen Sumpfsieben führen, während lösliche Korrosionsprodukte nicht zurückgehalten werden und somit in den Kern gelangen, was unter Umständen in Ausfällungsprozessen durch Temperaturänderungen resultiert. Da ein Einfluss dieser Effekte auf die Kernkühlung nicht ausgeschlossen werden kann, ist die Untersuchung der zugrundeliegenden physikochemischen Korrosions-, Ausfällungs- und Ablagerungsprozesse sowie deren thermohydraulischen Folgen Gegenstand von gemeinsamen Forschungsvorhaben des Helmholtz-Zentrums Dresden-Rossendorf, der TU Dresden sowie der Hochschule Zittau-Görlitz. Der Vortrag gibt einen Überblick über die bisherigen Forschungsarbeiten der o.g. Institutionen sowie die wesentlichen Ergebnisse der entsprechenden BMWi-Forschungsvorhaben im Kontext der Reaktorsicherheitsforschung.

Innerhalb des Vorhabens wurde ein ATHLET-CD-Eingabedatensatz für einen generischen deutschen DWR vom Typ KONVOI entwickelt. Das ATHLET-CD-Modell wurde für die Simulation schwerer Störfälle aus den Störfallkategorien Station Blackout (SBO) und Kühlmittelverluststörfällen mit kleinen Lecks (SBLOCA) eingesetzt. Dabei ist die vollständige Störfalltransiente für den Zeitbereich zwischen dem einleitenden Ereignis bis zum Versagen des Reaktordruckbehälters (RDB) abgedeckt und alle wesentli-chen Phänomene schwerer Störfällen werden abgebildet: Beginn der Kernaufheizung, Spaltproduktfrei-setzung, Aufschmelzen von Brennstoff- und Absorbermaterialien, Oxidationsprozesse mit Freisetzung von Wasserstoff, Verlagerung von geschmolzenem Material, Verlagerung in das untere Plenum, Schä-digung und Versagen des RDB. Das Modell wurde für die Analyse möglicher präventiver und mitigativer Notfallmaßnahmen für SBO und SBLOCA angewandt. Dafür wurden die Notfallmaßnahmen primärseitige Druckentlastung (PDE), primärseitiges Einspeisen mit mobilen Pumpensystemen sowie für SBLOCA das verzögerte Einspeisen der kaltseitigen Druckspeicher untersucht und die Eigenschaften und Einleitekriterien der Maßnahmen variiert. Es wurden die Zeitverläufe der Unfallszenarien analysiert und die verbleibenden Zeitspannen für die Einleitung zusätzlicher Maßnahmen ermittelt. Für ein SBO-Szenario mit PDE wurde für die Frühphase der Transiente (bis zum Beginn der Kernschmelze) eine Unsicherheits- und Sensititvitätsanalyse durchgeführt. Zusätzlich wurde für ein SBLOCA-Szenario ein Code-zu-Code-Vergleich zwischen ATHLET-CD und dem Störfallcode MELCOR erarbeitet.

In a joint research project of the Zittau/Goerlitz University of Applied Sciences (HSZG), the Technische Universitaet Dresden (TUD) and the Helmholtz-Zentrum Dresden-Rossendorf (HZDR), the main emphasis is the time-related assignment of simultaneous and interacting mechanisms at zinc sources and zinc sinks at boundary conditions of a loss-of-coolant accident (LOCA) in German pressurized water reactors (PWR). The according experiments are carried out at semi-technical as well as at laboratory scale. Zinc is used as a protective coating, e.g. for gratings in the containment, showing high corrosion resistance due to a gradual formation of passivating layers. In contrast, its long-term behaviour during LOCA changes significantly under the influence of the coolant chemistry of German PWR. As a consequence, according installations in the containment act as zinc sources due to corrosion. Released zinc ions change the chemical properties of the coolant and could e.g. lead to layer-forming depositions of zinc borates in the core, which increase the possibility of a hindered heat dissipation. For experimental and methodical investigations of these phenomena, the test rig “Zittau Flow Tray” (ZFT), a scaled sump model of a German PWR, was equipped with a full-length 3×3 fuel assembly (FA) dummy acting as core model, a preheater and a cooler component. Nine 4.4 m long fuel rod dummies simulate the decay heat by internal heating cartridges. This rig design enables experimental investigation of physico-chemical mechanisms considering coolant containing boric acid and zinc and their influence on the thermo-hydraulic processes in the reactor core at post-LOCA boundary conditions. The time depending zinc release at hot-dip galvanized gratings (HGG) was investigated regarding their position (e.g. inside or near the leaking jet, freely suspended or submerged in the coolant) and their surface area as well as temperature and flow rate of the coolant. The experimental database allows the

Innerhalb des Vorhabens wurde ein ATHLET-CD-Eingabedatensatz für einen generischen deutschen DWR vom Typ KONVOI entwickelt. Das ATHLET-CD-Modell wurde für die Simulation schwerer Störfälle aus den Störfallkategorien Station Blackout (SBO) und Kühlmittelverluststörfällen mit kleinen Lecks (SBLOCA) eingesetzt. Dabei ist die vollständige Störfalltransiente für den Zeitbereich zwischen dem einleitenden Ereignis bis zum Versagen des Reaktordruckbehälters (RDB) abgedeckt und alle wesentli-chen Phänomene schwerer Störfällen werden abgebildet: Beginn der Kernaufheizung, Spaltproduktfrei-setzung, Aufschmelzen von Brennstoff- und Absorbermaterialien, Oxidationsprozesse mit Freisetzung von Wasserstoff, Verlagerung von geschmolzenem Material, Verlagerung in das untere Plenum, Schä-digung und Versagen des RDB. Das Modell wurde für die Analyse möglicher präventiver und mitigativer Notfallmaßnahmen für SBO und SBLOCA angewandt. Dafür wurden die Notfallmaßnahmen primärseitige Druckentlastung (PDE), primärseitiges Einspeisen mit mobilen Pumpensystemen sowie für SBLOCA das verzögerte Einspeisen der kaltseitigen Druckspeicher untersucht und die Eigenschaften und Einleitekriterien der Maßnahmen variiert. Es wurden die Zeitverläufe der Unfallszenarien analysiert und die verbleibenden Zeitspannen für die Einleitung zusätzlicher Maßnahmen ermittelt. Für ein SBO-Szenario mit PDE wurde für die Frühphase der Transiente (bis zum Beginn der Kernschmelze) eine Unsicherheits- und Sensititvitätsanalyse durchgeführt. Zusätzlich wurde für ein SBLOCA-Szenario ein Code-zu-Code-Vergleich zwischen ATHLET-CD und dem Störfallcode MELCOR erarbeitet.

A historical review of in-vessel melt retention (IVR) is given, which is a severe accident mitigation measure extensively applied in Generation III pressurized water reactors (PWRs). The idea of IVR actually originated from the back-fitting of the Generation II reactor Loviisa VVER-440 in order to cope with the core-melt risk. It was then employed in the new deigns such as Westinghouse AP1000, the Korean APR1400 as well as Chinese advanced PWR designs HPR1000 and CAP1400. The most influential phenomena on the IVR strategy are in-vessel core melt evolution, the heat fluxes imposed on the vessel by the molten core, and the external cooling of the reactor pressure vessel (RPV). For in-vessel melt evolution, past focus has only been placed on the melt pool convection in the lower plenum of the RPV; however, through our review and analysis, we believe that other in-vessel phenomena, including core degradation and relocation, debris formation, and coolability and melt pool formation, may all contribute to the final state of the melt pool and its thermal loads on the lower head. By looking into previous research on relevant topics, we aim to identify the missing pieces in the picture. Based on the state of the art, we conclude by proposing future research needs., QC 20170102

Fully ceramic microencapsulated (FCM) fuel consists of TRISO (tristructural-isotropic) fuel particles embedded in a ceramic matrix (SiC) to form fuel pellets and rods and offers improved fission product retention and lower operating temperature with expected superior performance in normal and off-normal conditions compared to conventional fuel. When coupled with SiC cladding, FCM fuel eliminates zirconium altogether and is expected to drastically reduce hydrogen generation during a beyond-design- basis accident. In order to be deployed in current or future pressurized water reactors (PWRs), FCM fuel must meet or exceed the neutronic performance of conventional fuel. Limited by low heavy metal loading, an FCM fuel assembly requires increased enrichment and large fuel rods to match the cycle length of a conventional fuel assembly. This study investigated the core design, neutronics, and thermal hydraulics of a PWR loaded with FCM fuel and sought to optimize the assembly design to minimize the enrichment required to reach fuel performance similar to that of conventional fuel. It was found that the implementation of FCMfuel in a 17 X 17 assembly requires close to 20% enrichment and large fuel rods. Such design performs comparably to conventional fuel (4.5% enrichment) in terms of cycle length, reactivity coefficients, intra-assembly power peaking factor, burnable poison penalty, and control rod worth but requires an increase of pumping power. A parametric analysis spanned a large design space varying fuel outer diameter and pitch-to- diameter ratio (P/D) and downselected two alternate assembly designs: 11 X 11 (1.65-cm outer diameter and 1.18 P/D) and 9 X 9 (2.12-cm outer diameter and 1.12 P/D). These designs meet the cycle length requirement with 18.6% and 16.2% enrichments, respectively, but feature a smaller minimum departure from nucleate boiling ratio (MDNBR) compared to a reference assembly. It was estimated that a slight increase in rod outer diameter increases M

Fully ceramic microencapsulated (FCM) fuel consists of TRISO (tristructural-isotropic) fuel particles embedded in a ceramic matrix (SiC) to form fuel pellets and rods and offers improved fission product retention and lower operating temperature with expected superior performance in normal and off-normal conditions compared to conventional fuel. When coupled with SiC cladding, FCM fuel eliminates zirconium altogether and is expected to drastically reduce hydrogen generation during a beyond-design- basis accident. In order to be deployed in current or future pressurized water reactors (PWRs), FCM fuel must meet or exceed the neutronic performance of conventional fuel. Limited by low heavy metal loading, an FCM fuel assembly requires increased enrichment and large fuel rods to match the cycle length of a conventional fuel assembly. This study investigated the core design, neutronics, and thermal hydraulics of a PWR loaded with FCM fuel and sought to optimize the assembly design to minimize the enrichment required to reach fuel performance similar to that of conventional fuel. It was found that the implementation of FCMfuel in a 17 X 17 assembly requires close to 20% enrichment and large fuel rods. Such design performs comparably to conventional fuel (4.5% enrichment) in terms of cycle length, reactivity coefficients, intra-assembly power peaking factor, burnable poison penalty, and control rod worth but requires an increase of pumping power. A parametric analysis spanned a large design space varying fuel outer diameter and pitch-to- diameter ratio (P/D) and downselected two alternate assembly designs: 11 X 11 (1.65-cm outer diameter and 1.18 P/D) and 9 X 9 (2.12-cm outer diameter and 1.12 P/D). These designs meet the cycle length requirement with 18.6% and 16.2% enrichments, respectively, but feature a smaller minimum departure from nucleate boiling ratio (MDNBR) compared to a reference assembly. It was estimated that a slight increase in rod outer diameter increases M

During the sump recirculation phase after loss-of-coolant accidents (LOCAs) in pressurized water reactors (PWRs), coolant spilling from the leak in the primary cooling circuit is collected in the reactor sump and recirculated to the reactor core by residual-heat removal pumps as part of the emergency core cooling system (ECCS). Lab-scale studies within previous research projects have shown that the long-term contact of the boric acid containing coolant with hot-dip galvanized steel containment internals may cause corrosion of the corresponding materials influencing the cooling water chemistry due to dissolution of the zinc coat. Generic experimental investigations regarding the solubility of Zn corrosion products in boric acid solutions resulted in a decreasing solubility with increasing temperature. Thus, precipitation of solid corrosion products (zinc borates) cannot be ruled out if zinc containing coolant is heated up due to its recirculation into hot zones. Consequently, generic corrosion experiments were carried out in a lab-scale corrosion test facility aiming at the development and test of water-chemical measures to prevent zinc corrosion and zinc borate precipitation in boric acid containing coolants. The experimental results showed a decreasing corrosion rate with increasing pH value of the coolant. Thus, the risk of zinc borate precipitation can be reduced by addition of alkalizing media to the coolant after a LOCA. However, by adding of a moderate amount of alkalizing media to enhance the coolant pH into the neutral region, the zinc borate precipitation rate can be reduced only to about one third but not fully prevented. Extensive suppression of zinc corrosion and zinc borate precipitation is only achievable from a coolant pH of 7.5 resulting in a lithium concentration of 125 ppm if LiOH is used as alkalizing additive. Furthermore, foaming of the coolant cannot be ruled out if the coolant pH is increased into the slightly alkaline region.

During the sump recirculation phase after loss-of-coolant accidents (LOCA) in pressurized water reactors (PWR), coolant spilling out of the leak in the primary cooling circuit is collected in the reactor sump and recirculated to the reactor core by residual-heat removal pumps as part of the emergency core cooling system (ECCS). The contact of the coolant with several forms of debris may influence the sump strainer clogging behavior as well as the cooling water chemistry. Damage to fibrous insulation materials located near to the leak may compromise the operation of the ECCS, if insulation fibers are transported to the strainers. Furthermore, the long-term contact of the boric acid containing coolant with hot-dip galvanized containment internals (e.g. grating treads, supporting grids of sump strainers) may cause corrosion of the corresponding materials. Generic investigations regarding the influence of such corrosion processes on strainer clogging as well as on the coolant chemistry and possible resulting in-core effects are subject of joint research projects of the Helmholtz-Zentrum Dresden-Rossendorf (HZDR), TU Dresden (TUD) and Zittau-Görlitz University of Applied Sciences (HSZG). Lab-scale experiments at HZDR and TUD are focused on elucidation of physico-chemical corrosion and precipitation processes as well as resulting clogging effects. Results of generic experiments in a lab-scale corrosion test facility, representing the ECCS operation in a simplified manner, suggest that there is a multi-stage corrosion process. The first stage comprises dissolution of the zinc layer in the coolant forming zinc ions and in turn affecting the coolant chemistry. During the second stage, the base material (steel) corrodes forming insoluble corrosion particles, which can subsequently lead to accelerated clogging of fiber-laden strainers within a few hours. The main influences on corrosion were identified as impact of the coolant jet onto the corroding surface, water chemistry an

During the sump recirculation phase after loss-of-coolant accidents (LOCAs) in pressurized water reactors (PWRs), coolant spilling from the leak in the primary cooling circuit is collected in the reactor sump and recirculated to the reactor core by residual-heat removal pumps as part of the emergency core cooling system (ECCS). Lab-scale studies within previous research projects have shown that the long-term contact of the boric acid containing coolant with hot-dip galvanized steel containment internals may cause corrosion of the corresponding materials influencing the cooling water chemistry due to dissolution of the zinc coat. Generic experimental investigations regarding the solubility of Zn corrosion products in boric acid solutions resulted in a decreasing solubility with increasing temperature. Thus, precipitation of solid corrosion products (zinc borates) cannot be ruled out if zinc containing coolant is heated up due to its recirculation into hot zones. Consequently, generic corrosion experiments were carried out in a lab-scale corrosion test facility aiming at the development and test of water-chemical measures to prevent zinc corrosion and zinc borate precipitation in boric acid containing coolants. The experimental results showed a decreasing corrosion rate with increasing pH value of the coolant. Thus, the risk of zinc borate precipitation can be reduced by addition of alkalizing media to the coolant after a LOCA. However, by adding of a moderate amount of alkalizing media to enhance the coolant pH into the neutral region, the zinc borate precipitation rate can be reduced only to about one third but not fully prevented. Extensive suppression of zinc corrosion and zinc borate precipitation is only achievable from a coolant pH of 7.5 resulting in a lithium concentration of 125 ppm if LiOH is used as alkalizing additive. Furthermore, foaming of the coolant cannot be ruled out if the coolant pH is increased into the slightly alkaline region.

During the sump recirculation phase after loss-of-coolant accidents (LOCA) in pressurized water reactors (PWR), coolant spilling out of the leak in the primary cooling circuit is collected in the reactor sump and recirculated to the reactor core by residual-heat removal pumps as part of the emergency core cooling system (ECCS). The contact of the coolant with several forms of debris may influence the sump strainer clogging behavior as well as the cooling water chemistry. Damage to fibrous insulation materials located near to the leak may compromise the operation of the ECCS, if insulation fibers are transported to the strainers. Furthermore, the long-term contact of the boric acid containing coolant with hot-dip galvanized containment internals (e.g. grating treads, supporting grids of sump strainers) may cause corrosion of the corresponding materials. Generic investigations regarding the influence of such corrosion processes on strainer clogging as well as on the coolant chemistry and possible resulting in-core effects are subject of joint research projects of the Helmholtz-Zentrum Dresden-Rossendorf (HZDR), TU Dresden (TUD) and Zittau-Görlitz University of Applied Sciences (HSZG). Lab-scale experiments at HZDR and TUD are focused on elucidation of physico-chemical corrosion and precipitation processes as well as resulting clogging effects. Results of generic experiments in a lab-scale corrosion test facility, representing the ECCS operation in a simplified manner, suggest that there is a multi-stage corrosion process. The first stage comprises dissolution of the zinc layer in the coolant forming zinc ions and in turn affecting the coolant chemistry. During the second stage, the base material (steel) corrodes forming insoluble corrosion particles, which can subsequently lead to accelerated clogging of fiber-laden strainers within a few hours. The main influences on corrosion were identified as impact of the coolant jet onto the corroding surface, water chemistry an

Within the framework of the German reactor safety research, generic experimental investigations were carried out aiming at thermal-hydraulic consequences of physicochemical mechanisms, caused by dissolution of zinc in boric acid during corrosion processes at hot-dip galvanized surfaces of containment internals at lower coolant temperatures and the subsequent precipitation of solid zinc borates in PWR core regions of higher temperature. This constellation can occur during sump recirculation operation of ECCS after LOCA. Hot-dip galvanized components, which are installed inside a PWR containment, may act as zinc sources. Getting in contact with boric acid coolant, zinc at their surfaces is released into coolant in form of ions due to corrosion processes. As a long-term behavior resp. over a time period of several days, metal layers of zinc and zinc alloys can dissolve extensively. First fundamental studies at laboratory scale were done at the Helmholtz-Zentrum Dresden-Rossendorf (HZDR). Their experimental results were picked up for the definition of boundary conditions for experiments at semi-technical scale at the University of Applied Sciences Zittau/Görlitz (HSZG). Electrical heating rods with zircaloy cladding tubes have been used as fuel rod simulators. As near-plant core components, a 3x3 configuration of heating rods (HRC) and a shortened, partially heatable PWR fuel assembly dummy were applied into cooling circuits. The HRC module include segments of spacers for a suitable representation of a heating channel geometry. Formations of different solid zinc compounds (mainly borates) were observed at heatable zircaloy surfaces and characterized in detail for heating-up to several coolant temperatures. As a strict consequence of their proven influence on heat removal and coolant flow behavior in the PWR core, preventive water-chemical methods were defined and tested.

Within the framework of the German reactor safety research, generic experimental investigations were carried out aiming at thermal-hydraulic consequences of physicochemical mechanisms, caused by dissolution of zinc in boric acid during corrosion processes at hot-dip galvanized surfaces of containment internals at lower coolant temperatures and the subsequent precipitation of solid zinc borates in PWR core regions of higher temperature. This constellation can occur during sump recirculation operation of ECCS after LOCA. Hot-dip galvanized components, which are installed inside a PWR containment, may act as zinc sources. Getting in contact with boric acid coolant, zinc at their surfaces is released into coolant in form of ions due to corrosion processes. As a long-term behavior resp. over a time period of several days, metal layers of zinc and zinc alloys can dissolve extensively. First fundamental studies at laboratory scale were done at the Helmholtz-Zentrum Dresden-Rossendorf (HZDR). Their experimental results were picked up for the definition of boundary conditions for experiments at semi-technical scale at the University of Applied Sciences Zittau/Görlitz (HSZG). Electrical heating rods with zircaloy cladding tubes have been used as fuel rod simulators. As near-plant core components, a 3x3 configuration of heating rods (HRC) and a shortened, partially heatable PWR fuel assembly dummy were applied into cooling circuits. The HRC module include segments of spacers for a suitable representation of a heating channel geometry. Formations of different solid zinc compounds (mainly borates) were observed at heatable zircaloy surfaces and characterized in detail for heating-up to several coolant temperatures. As a strict consequence of their proven influence on heat removal and coolant flow behavior in the PWR core, preventive water-chemical methods were defined and tested.

During the sump recirculation phase after loss-of-coolant accidents (LOCA) in pressurized water reactors (PWR), coolant spilling out of the leak in the primary cooling circuit is collected in the reactor sump and recirculated to the reactor core by residual-heat removal pumps as part of the emergency core cooling system (ECCS). The contact of the coolant with several forms of debris may influence the sump strainer clogging behavior as well as the cooling water chemistry. Damage to fibrous insulation materials located near to the leak may compromise the operation of the ECCS, if insulation fibers are transported to the strainers. Furthermore, the long-term contact of the boric acid containing coolant with hot-dip galvanized containment internals (e.g. grating treads, supporting grids of sump strainers) may cause corrosion of the corresponding materials. Generic investigations regarding the influence of such corrosion processes on strainer clogging as well as on the coolant chemistry and possible resulting in-core effects are subject of joint research projects of the Helmholtz-Zentrum Dresden-Rossendorf (HZDR), TU Dresden (TUD) and Zittau-Görlitz University of Applied Sciences (HSZG). Lab-scale experiments at HZDR and TUD are focused on elucidation of physico-chemical corrosion and precipitation processes as well as resulting clogging effects. Results of generic experiments in a lab-scale corrosion test facility, representing the ECCS operation in a simplified manner, suggest that there is a multi-stage corrosion process. The first stage comprises dissolution of the zinc layer in the coolant forming zinc ions and in turn affecting the coolant chemistry. During the second stage, the base material (steel) corrodes forming insoluble corrosion particles, which can subsequently lead to accelerated clogging of fiber-laden strainers within a few hours. The main influences on corrosion were identified as impact of the coolant jet onto the corroding surface, water chemistry an

During the sump recirculation phase after loss-of-coolant accidents (LOCAs) in pressurized water reactors (PWRs), coolant spilling from the leak in the primary cooling circuit is collected in the reactor sump and recirculated to the reactor core by residual-heat removal pumps as part of the emergency core cooling system (ECCS). Lab-scale studies within previous research projects have shown that the long-term contact of the boric acid containing coolant with hot-dip galvanized steel containment internals may cause corrosion of the corresponding materials influencing the cooling water chemistry due to dissolution of the zinc coat. Generic experimental investigations regarding the solubility of Zn corrosion products in boric acid solutions resulted in a decreasing solubility with increasing temperature. Thus, precipitation of solid corrosion products (zinc borates) cannot be ruled out if zinc containing coolant is heated up due to its recirculation into hot zones. Consequently, generic corrosion experiments were carried out in a lab-scale corrosion test facility aiming at the development and test of water-chemical measures to prevent zinc corrosion and zinc borate precipitation in boric acid containing coolants. The experimental results showed a decreasing corrosion rate with increasing pH value of the coolant. Thus, the risk of zinc borate precipitation can be reduced by addition of alkalizing media to the coolant after a LOCA. However, by adding of a moderate amount of alkalizing media to enhance the coolant pH into the neutral region, the zinc borate precipitation rate can be reduced only to about one third but not fully prevented. Extensive suppression of zinc corrosion and zinc borate precipitation is only achievable from a coolant pH of 7.5 resulting in a lithium concentration of 125 ppm if LiOH is used as alkalizing additive. Furthermore, foaming of the coolant cannot be ruled out if the coolant pH is increased into the slightly alkaline region.

During the sump recirculation phase after loss-of-coolant accidents (LOCA) in pressurized water reactors (PWR), coolant spilling out of the leak in the primary cooling circuit is collected in the reactor sump and recirculated to the reactor core by residual-heat removal pumps as part of the emergency core cooling system (ECCS). The contact of the coolant with several forms of debris may influence the sump strainer clogging behavior as well as the cooling water chemistry. Damage to fibrous insulation materials located near to the leak may compromise the operation of the ECCS, if insulation fibers are transported to the strainers. Furthermore, the long-term contact of the boric acid containing coolant with hot-dip galvanized containment internals (e.g. grating treads, supporting grids of sump strainers) may cause corrosion of the corresponding materials. Generic investigations regarding the influence of such corrosion processes on strainer clogging as well as on the coolant chemistry and possible resulting in-core effects are subject of joint research projects of the Helmholtz-Zentrum Dresden-Rossendorf (HZDR), TU Dresden (TUD) and Zittau-Görlitz University of Applied Sciences (HSZG). Lab-scale experiments at HZDR and TUD are focused on elucidation of physico-chemical corrosion and precipitation processes as well as resulting clogging effects. Results of generic experiments in a lab-scale corrosion test facility, representing the ECCS operation in a simplified manner, suggest that there is a multi-stage corrosion process. The first stage comprises dissolution of the zinc layer in the coolant forming zinc ions and in turn affecting the coolant chemistry. During the second stage, the base material (steel) corrodes forming insoluble corrosion particles, which can subsequently lead to accelerated clogging of fiber-laden strainers within a few hours. The main influences on corrosion were identified as impact of the coolant jet onto the corroding surface, water chemistry an

During the sump recirculation phase after loss-of-coolant accidents (LOCAs) in pressurized water reactors (PWRs), coolant spilling from the leak in the primary cooling circuit is collected in the reactor sump and recirculated to the reactor core by residual-heat removal pumps as part of the emergency core cooling system (ECCS). Lab-scale studies within previous research projects have shown that the long-term contact of the boric acid containing coolant with hot-dip galvanized steel containment internals may cause corrosion of the corresponding materials influencing the cooling water chemistry due to dissolution of the zinc coat. Generic experimental investigations regarding the solubility of Zn corrosion products in boric acid solutions resulted in a decreasing solubility with increasing temperature. Thus, precipitation of solid corrosion products (zinc borates) cannot be ruled out if zinc containing coolant is heated up due to its recirculation into hot zones. Consequently, generic corrosion experiments were carried out in a lab-scale corrosion test facility aiming at the development and test of water-chemical measures to prevent zinc corrosion and zinc borate precipitation in boric acid containing coolants. The experimental results showed a decreasing corrosion rate with increasing pH value of the coolant. Thus, the risk of zinc borate precipitation can be reduced by addition of alkalizing media to the coolant after a LOCA. However, by adding of a moderate amount of alkalizing media to enhance the coolant pH into the neutral region, the zinc borate precipitation rate can be reduced only to about one third but not fully prevented. Extensive suppression of zinc corrosion and zinc borate precipitation is only achievable from a coolant pH of 7.5 resulting in a lithium concentration of 125 ppm if LiOH is used as alkalizing additive. Furthermore, foaming of the coolant cannot be ruled out if the coolant pH is increased into the slightly alkaline region.

In the case of a severe reactor accident, knowledge of the coolant level and the state of the core, including the progress of a possible core melt, would be of crucial interest. In such a scenario, the in-core instrumentation will most likely not be available. In this work, we explore the possibility of using the ex-core neutron detectors to gain information about the state of the reactor pressure vessel (RPV) inventory for a light water reactor. These detectors, which are typically implemented as ionization chambers, are located inside the biological shield and might still be operational during a severe accident. Stationary Monte Carlo calculations using the radiation transport code MCNP were performed to simulate the transport of neutrons outside the RPV and the reactions of the ionization chambers. These detector signals are computed for different model reactor states which might occur during a severe accident with core meltdown. The reactor model is based on data from a typical German Pressurized Water Reactor. The results indicate that a change in coolant level should be detectable. Due to the core’s neutron self-shielding, deformations in the inner core region, such as the formation of a cavity, do not yield different signal rates in the ex-core neutron chambers. Changes not confined to the centre of the core, such as the relocation of corium into the lower head, re detectable by their change in the ionization chambers’ reaction rates.

In the case of a severe reactor accident, knowledge of the coolant level and the state of the core, including the progress of a possible core melt, would be of crucial interest. In such a scenario, the in-core instrumentation will most likely not be available. In this work, we explore the possibility of using the ex-core neutron detectors to gain information about the state of the reactor pressure vessel (RPV) inventory for a light water reactor. These detectors, which are typically implemented as ionization chambers, are located inside the biological shield and might still be operational during a severe accident. Stationary Monte Carlo calculations using the radiation transport code MCNP were performed to simulate the transport of neutrons outside the RPV and the reactions of the ionization chambers. These detector signals are computed for different model reactor states which might occur during a severe accident with core meltdown. The reactor model is based on data from a typical German Pressurized Water Reactor. The results indicate that a change in coolant level should be detectable. Due to the core’s neutron self-shielding, deformations in the inner core region, such as the formation of a cavity, do not yield different signal rates in the ex-core neutron chambers. Changes not confined to the centre of the core, such as the relocation of corium into the lower head, re detectable by their change in the ionization chambers’ reaction rates.

In the case of a severe reactor accident, knowledge of the coolant level and the state of the core, including the progress of a possible core melt, would be of crucial interest. In such a scenario, the in-core instrumentation will most likely not be available. In this work, we explore the possibility of using the ex-core neutron detectors to gain information about the state of the reactor pressure vessel (RPV) inventory for a light water reactor. These detectors, which are typically implemented as ionization chambers, are located inside the biological shield and might still be operational during a severe accident. Stationary Monte Carlo calculations using the radiation transport code MCNP were performed to simulate the transport of neutrons outside the RPV and the reactions of the ionization chambers. These detector signals are computed for different model reactor states which might occur during a severe accident with core meltdown. The reactor model is based on data from a typical German Pressurized Water Reactor. The results indicate that a change in coolant level should be detectable. Due to the core’s neutron self-shielding, deformations in the inner core region, such as the formation of a cavity, do not yield different signal rates in the ex-core neutron chambers. Changes not confined to the centre of the core, such as the relocation of corium into the lower head, re detectable by their change in the ionization chambers’ reaction rates.

In the case of a severe reactor accident, knowledge of the coolant level and the state of the core, including the progress of a possible core melt, would be of crucial interest. In such a scenario, the in-core instrumentation will most likely not be available. In this work, we explore the possibility of using the ex-core neutron detectors to gain information about the state of the reactor pressure vessel (RPV) inventory for a light water reactor. These detectors, which are typically implemented as ionization chambers, are located inside the biological shield and might still be operational during a severe accident. Stationary Monte Carlo calculations using the radiation transport code MCNP were performed to simulate the transport of neutrons outside the RPV and the reactions of the ionization chambers. These detector signals are computed for different model reactor states which might occur during a severe accident with core meltdown. The reactor model is based on data from a typical German Pressurized Water Reactor. The results indicate that a change in coolant level should be detectable. Due to the core’s neutron self-shielding, deformations in the inner core region, such as the formation of a cavity, do not yield different signal rates in the ex-core neutron chambers. Changes not confined to the centre of the core, such as the relocation of corium into the lower head, re detectable by their change in the ionization chambers’ reaction rates.

Within the framework of German nuclear safety research, generic experimental investigations were carried out at HZDR and the Hochschule Zittau/Görlitz aiming at the elucidation of physicochemical and thermo hydraulic mechanism of corrosion product formation, which may occur during the sump circulation operation after loss-of-coolant accidents in pressurized water reactors. The contact of the boric acid containing coolant with hot-dip galvanized steel containment internals causes corrosion of the corresponding materials resulting in dissolution of the zinc coating. As main result of batch experiments, decreasing solubility of zinc corrosion products in boric acid solutions with increasing temperature was found. Thus, the formation and deposition of solid corrosion products cannot be ruled out if zinc containing coolant is heated up due to its recirculation into hot regions within the cooling circuit. Generic corrosion and deposition experiments at a lab-scale test facility proved that dissolved zinc, formed at low temperatures in boric acid solution by zinc corrosion, may turn into solid deposits of zinc borates when contacting heated zircaloy surfaces during the heating of the coolant. The results obtained at lab-scale were confirmed by generic experiments at semi-technical scale using a 3x3 heating rod configuration including spacer segments as well as a 16x16 (8x8 heated) fuel rod dummy. Experiments regarding the corrosion kinetics of zinc-coated components were not subject of the study. Therefore, a quantitative transferability of the results to postulated PWR-LOCA is not given so far.

During loss-of-coolant accidents (LOCAs) in pressurized water reactors (PWRs), coolant spilling from the leak in the primary cooling circuit is collected in the reactor sump and recirculated to the reactor core by low pressure injection pumps as part of the emergency core cooling system. The long-term contact of the boric acid containing coolant with hot-dip galvanized steel containment inter-nals (e.g. grating treads, channels, supporting grids of sump strainers) may cause corrosion of the corresponding materials influencing the cooling water chemistry due to dissolution of the zinc coating. Experimental investigations regarding the solubility of Zn corrosion products in boric acid solutions resulted in a decreasing solubility with increasing temperature. Thus, the formation of solid (i.e. particulate) corrosion products cannot be ruled out if the Zn containing coolant is heated up due to its recirculation into hot zones. During lab-scale experiments, significant amounts of solid corrosion products have been found as deposited layers on hot surfaces as well as in the form of deposits at tubes, fittings and retaining components depending on formation temperature and hydrodynamic conditions. The solid corrosion products were identified as zinc borates. Depending on their forming temperature, different zinc borate compounds may occur having different physicochemical properties. Although the kinetics of the processes obtained at lab-scale are not transferable to those proceeding during a PWR LOCA due to their dependency on the corroding surface area as well as on the local thermal hydraulics, the results give an insight into physicochemical processes, which might occur in case of zinc corrosion in cooling circuits.

Im Rahmen der Reaktorsicherheitsforschung erfolgten an der Hochschule Zittau/Görlitz in Kooperation mit dem Helmholtz-Zentrum Dresden-Rossendorf experimentelle und methodische Untersuchungen für die systematische Klärung physiko-chemischer Mechanismen und deren Auswirkungen auf thermo-fluiddynamische Prozesse, welche während des Sumpfumwälzbetriebs nach Kühlmittelverluststörfällen in einem Kernkraftwerk ablaufen können, falls in boriertem Kühlmittel (KM) gelöstes Zink in Kernbereiche höherer Temperatur (Hot-Spots) gelangt. Das im KM befindliche Zink kann hierbei im Vorfeld durch die Korrosion feuerverzinkter Bauteile freigesetzt werden. In den Untersuchungen im halbtechnischen Maßstab wurden die physiko-chemischen Mechanismen und der Temperatureinfluss analysiert. Gleichzeitig wurden Auswirkungen dieser Prozesse auf das thermo-fluiddynamische Verhalten in einer Heizstabkonfiguration (3x3-Anordnung mit für Druckwasserreaktoren (DWR) typischen Zirkaloy-Hüllrohren) mit Abstandshaltern erfasst. Im Fokus der Untersuchungen stand dabei das Verhalten derart zusammengesetzter Fluide an beheizten Konfigurationen, die im Kern von DWR auftreten können. Die durchgeführten Untersuchungen tragen generischen Charakter und liefern Aussagen zum Löslichkeitsverhalten von Zink in borsäurehaltigem KM sowie zur Bildung fester Korrosionsprodukte und den daraus folgenden Auswirkungen.

During loss-of-coolant accidents (LOCAs) in pressurized water reactors (PWRs), coolant spilling from the leak in the primary cooling circuit is collected in the reactor sump and recirculated to the reactor core by low pressure injection pumps as part of the emergency core cooling system. The long-term contact of the boric acid containing coolant with hot-dip galvanized steel containment inter-nals (e.g. grating treads, channels, supporting grids of sump strainers) may cause corrosion of the corresponding materials influencing the cooling water chemistry due to dissolution of the zinc coating. Experimental investigations regarding the solubility of Zn corrosion products in boric acid solutions resulted in a decreasing solubility with increasing temperature. Thus, the formation of solid (i.e. particulate) corrosion products cannot be ruled out if the Zn containing coolant is heated up due to its recirculation into hot zones. During lab-scale experiments, significant amounts of solid corrosion products have been found as deposited layers on hot surfaces as well as in the form of deposits at tubes, fittings and retaining components depending on formation temperature and hydrodynamic conditions. The solid corrosion products were identified as zinc borates. Depending on their forming temperature, different zinc borate compounds may occur having different physicochemical properties. Although the kinetics of the processes obtained at lab-scale are not transferable to those proceeding during a PWR LOCA due to their dependency on the corroding surface area as well as on the local thermal hydraulics, the results give an insight into physicochemical processes, which might occur in case of zinc corrosion in cooling circuits.

Within the framework of the reactor safety research, generic experimental investigations were carried out aiming at the physicochemical background of possible zinc corrosion product formation, which may occur inside the reactor pressure vessel during the sump circulation operation after loss-of-coolant accidents in pressurized water reactors. The contact of the boric acid containing coolant with hot-dip galvanized steel containment internals causes corrosion of the corresponding materials resulting in dissolution of the zinc coat. A retrograde solubility of zinc corrosion products with increasing temperature was observed during batch experiments of zinc corrosion in boric acid containing coolants. Thus, the formation and deposition of solid corrosion products cannot be ruled out if the coolant containing dissolved zinc is heated up during its recirculation into hot regions within the emergency cooling circuit (e.g. hot-spots in the core). Corrosion experiments at a lab-scale test facility, which included formation of corrosion products at a single heated cladding tube, proved that dissolved zinc, formed at low temperatures in boric acid solution by zinc corrosion, turns into solid deposits of zinc borates when contacting heated zircaloy surfaces during the heating of the coolant. Moreover, the temperature of formation influences the chemical composition of the zinc borates and thus the deposition and mobilization behavior of the products.

During loss-of-coolant accidents (LOCAs) in pressurized water reactors (PWRs), coolant spilling from the leak in the primary cooling circuit is collected in the reactor sump and recirculated to the reactor core by low pressure injection pumps as part of the emergency core cooling system. The long-term contact of the boric acid containing coolant with hot-dip galvanized steel containment inter-nals (e.g. grating treads, channels, supporting grids of sump strainers) may cause corrosion of the corresponding materials influencing the cooling water chemistry due to dissolution of the zinc coating. Experimental investigations regarding the solubility of Zn corrosion products in boric acid solutions resulted in a decreasing solubility with increasing temperature. Thus, the formation of solid (i.e. particulate) corrosion products cannot be ruled out if the Zn containing coolant is heated up due to its recirculation into hot zones. During lab-scale experiments, significant amounts of solid corrosion products have been found as deposited layers on hot surfaces as well as in the form of deposits at tubes, fittings and retaining components depending on formation temperature and hydrodynamic conditions. The solid corrosion products were identified as zinc borates. Depending on their forming temperature, different zinc borate compounds may occur having different physicochemical properties. Although the kinetics of the processes obtained at lab-scale are not transferable to those proceeding during a PWR LOCA due to their dependency on the corroding surface area as well as on the local thermal hydraulics, the results give an insight into physicochemical processes, which might occur in case of zinc corrosion in cooling circuits.

Im Rahmen der Reaktorsicherheitsforschung erfolgten an der Hochschule Zittau/Görlitz in Kooperation mit dem Helmholtz-Zentrum Dresden-Rossendorf experimentelle und methodische Untersuchungen für die systematische Klärung physiko-chemischer Mechanismen und deren Auswirkungen auf thermo-fluiddynamische Prozesse, welche während des Sumpfumwälzbetriebs nach Kühlmittelverluststörfällen in einem Kernkraftwerk ablaufen können, falls in boriertem Kühlmittel (KM) gelöstes Zink in Kernbereiche höherer Temperatur (Hot-Spots) gelangt. Das im KM befindliche Zink kann hierbei im Vorfeld durch die Korrosion feuerverzinkter Bauteile freigesetzt werden. In den Untersuchungen im halbtechnischen Maßstab wurden die physiko-chemischen Mechanismen und der Temperatureinfluss analysiert. Gleichzeitig wurden Auswirkungen dieser Prozesse auf das thermo-fluiddynamische Verhalten in einer Heizstabkonfiguration (3x3-Anordnung mit für Druckwasserreaktoren (DWR) typischen Zirkaloy-Hüllrohren) mit Abstandshaltern erfasst. Im Fokus der Untersuchungen stand dabei das Verhalten derart zusammengesetzter Fluide an beheizten Konfigurationen, die im Kern von DWR auftreten können. Die durchgeführten Untersuchungen tragen generischen Charakter und liefern Aussagen zum Löslichkeitsverhalten von Zink in borsäurehaltigem KM sowie zur Bildung fester Korrosionsprodukte und den daraus folgenden Auswirkungen.