A single fuel channel break is the accident when the fuel channel is overheated during the reactor operation on power at nominal pressure. In case of other accidents that are included in this group (1.2, see Figure 5), the overheating of fuel can occur after the reactor shutdown at low pressure in RCS, because the strength of fuel channels at nominal pressure at the temperature margin of 650-800 oC is limited [4]. Otherwise (if FC walls temperature exceeds this limit in few fuel channels during normal pressure in the circuit) multiple ruptures of fuel channels can occur. As design basis accident for RC is a single FC rupture, the accident with
multiple ruptures of fuel channels is included into the group of accidents when the reactor core is completely damaged and the integrity of the reactor constructions is not preserved. The reactor cavity venting system at Ignalina NPP with RBMK-1500 reactor was improved providing the additional flow path to ALS, i. e. the RCVS capacity was increased to withstand a multiple rupture of fuel channels. Figure 18 presents the summary of the analysis performed to estimate the number of fuel channels that can be ruptured simultaneously in the beginning of the accident, i. e. at nominal RCS pressure and temperature, and reactor power of 4200 MWth, the integrity of the reactor cavity would be maintained. A detailed analysis is presented in [21]. The coolant release rate was calculated using code RELAP5 (model presented in Figure 6) and the analysis of the Reactor Cavity and ALS response was performed using code CONTAIN. The performed analysis showed that making the most conservative assumptions, the reactor cavity could withstand simultaneous rupture of at least 11-16 FCs.

0 5 10 15 20

Number of ruptured FC

1.1.1 Design

The purpose of designing a reduced-scale model system for the modularization of the reactor internals is to confirm the performance and application conditions of a remote distance measurement sensor selected. The sensor should be able to measure gaps between the CSB snubber lug and the RV core-stabilizing lug in the range 0.381-0.508 [mm].

This system was designed to be used at narrow space and small probe hole conditions and to be light easy to handle by using aluminum material. A threaded connection jig allows the remote measurement sensor to be assembled into the CSB snubber.

The block diagram in Fig. 5 shows the reduced-scale model system. The servomotors provide the movement in the up/ down and front/rear directions for the reduced-scale models of the CSB snubber lug with remote measurement sensors connected. The air compressor drives the measurement sensors and the obtained data are displayed on the computer screen through the interface modules and network cables.

The picture in Fig. 6 shows the reduced-scale model system that consists of the reduced — scale CSB snubber lug, the reduced-scale RV core-stabilizing lug, 4 remote distance measurement sensors, the air supply device, 2 servo motor devices, the data interface modules and a laptop computer.

The components of the reduced-scale model system were fabricated as follows.

— The model pieces of the CSB snubber lug and RV lug were machined to precisely represent the surface and inside of the remote distance measurement sensors’ hole. The material is aluminum alloy.

— The threaded connection jig of the remote distance measurement sensor was fabricated.

— The zero point adjustment device was fabricated.

— The block gauge was used to test the reliability of the remote distance measurement sensors.

(a)

‘ensor head

Reduced-scale

Gap

(b)

(a) Overall system, (b) Enlarged circle of Fig. 6 (a) to simulate measurement of gaps between CSB snubber lug and RV core-stabilizing lug

Fulvio Mascari1, Giuseppe Vella1, Brian G. Woods2, Kent Welter3 and Francesco D’Auria4

1Department of Energy, The University of Palermo 2Department of Nuclear Engineering and Radiation Health Physics

Oregon State University 3NuScale Power Inc.

4San Piero a Grado — Nuclear Research Group (SPGNRG), University of Pisa

1,4Italy

2,3USA

1. Introduction

Today considering the world energy demand increase, the use of advanced nuclear power plants, have an important role in the environment and economic sustainability of country energy strategy mix considering the capacity of nuclear reactors of producing energy in safe and stable way contributing in cutting the CO2 emission (Bertel & Morrison, 2001; World Energy Outlook-Executive Summary, 2009; Wolde-Rufael & Menyah, 2010; Mascari et al., 2011d). According to the information’s provided by the "Power Reactor Information System" of the International Atomic Energy Agency (IAEA), today 433 nuclear power reactors are in operation in the world providing a total power installed capacity of 366.610 GWe, 5 nuclear reactors are in long term shutdown and 65 units are under construction (IAEA PRIS, 2011).

In the last 20 years, the international community, taking into account the operational experience of the nuclear reactors, starts the development of new advanced reactor designs, to satisfy the demands of the people to improve the safety of nuclear power plants and the demands of the utilities to improve the economic efficiency and reduce the capital costs (D’Auria et al., 1993; Mascari et al., 2011c). Design simplifications and increased design margins are included in the advanced Light Water Reactors (LWR) (Aksan, 2005). In this framework, the project of some advanced reactors considers the use of emergency systems based entirely on natural circulation for the removal of the decay power in transient condition and in some reactors for the removal of core power during normal operating conditions (IAEA-TECDOC-1624, 2009; Mascari et al., 2010a; Mascari et al., 2011d). For example, if the normal heat sink is not available, the decay heat can be removed by using a passive connection between the primary system and heat exchangers (Aksan, 2005; Mascari et al., 2010a, Mascari, 2010b). The AP600/1000 (Advanced Plant 600/1000 MWe) design, for

example, includes a Passive Residual Heat Removal (PRHR) system consisting of a C-Tube type heat exchanger immersed in the In-containment Refueling Water Storage Tank (IRWST) and connected to one of the Hot Legs (HL) (IAEA-TECDOC-1391, 2004; Reyes, 2005c; Gou et al., 2009; Mascari et al., 2010a). A PRHR from the core via Steam Generators (SG) to the atmosphere, considered in the WWER-1000/V-392 (Water Moderated, Water Cooled Energy Reactor) design, consists of heat exchangers cooled by atmospheric air, while the PRHR via SGs, considered in the WWER-640/V-407 design, consists of heat exchangers immersed in emergency heat removal tanks installed outside the containment (Kurakov et al., 2002; IAEA-TECDOC-1391, 2004; Gou et al., 2009; Mascari et al., 2010a). In the AC-600 (Advanced Chinese PWR) the PRHR heat exchangers are cooled by atmospheric air (IAEA — TECDOC 1281, 2002; Zejun et al., 2003; IAEA-TECDOC-1391, 2004; Gou et al., 2009; Mascari et al., 2010a) and in the System Integrated Modular Advanced Reactor (SMART) the PRHR heat exchangers are submerged in an in-containment refuelling water tank (IAEA-TECDOC — 1391, 2004; Lee & Kim, 2008; Gou et al., 2009; Mascari et al., 2010a). The International Reactor Innovative and Secure (IRIS) design includes a passive Emergency Heat Removal System (EHRS) consisting of an heat exchanger immersed in the Refueling Water Storage Tank (RWST). The EHRS is connected to a separate SG feed and steam line and the RWST is installed outside the containment structure (Carelli et al., 2004; Carelli et al., 2009; Mascari, 2010b; Chiovaro et al., 2011). In the advanced BWR designs the core water evaporates, removing the core decay heat, and condenses in a heat exchanger placed in a pool. Then the condensate comes back to the core (Hicken & Jaegers, 2002; Mascari et al., 2010a). For example, the SWR-1000 (Siede Wasser Reaktor, 1000 MWe) design has emergency condensers immersed in a core flooding pool and connected to the core, while the ESBWR (Economic Simplified Boiling Water Reactor) design uses isolation condensers connected to the Reactor Pressure Vessel (RPV) and immersed in external pools (IAEA-TECDOC-1391, 2004; Aksan, 2005; Mascari et al., 2010a).

The designs of some advanced reactors rely on natural circulation for the removing of the core power during normal operation. Examples of these reactors are the MASLWR (Multi­Application Small Light Water Reactor), the ESBWR, the SMART and the Natural Circulation based PWR being developed in Argentina (CAREM)(IAEA-TECDOC-1391, 2004; IAEA -TECDOC-1474, 2005; Mascari et al., 2010a). In particular the MASLWR (Modro et al., 2003), figure 1, is a small modular integral Pressurized Water Reactor (PWR) relying on natural circulation during both steady-state and transient operation.

In the development process of these advanced nuclear reactors, the analysis of single and two-phase fluid natural circulation in complex systems (Zuber, 1991; Levy, 1999; Reyes & King, 2003; IAEA-TECDOC-1474, 2005; Mascari et al., 2011e), under steady state and transient conditions, is crucial for the understanding of the physical and operational phenomena typical of these advanced designs. The use of experimental facilities is fundamental in order to characterize the thermal hydraulics of these phenomena and to develop an experimental database useful for the validation of the computational tools necessary for the operation, design and safety analysis of nuclear reactors. In general it is expensive to design a test facility to develop experimental data useful for the analyses of complex system, therefore reduced scaled test facilities are, in general, used to characterize them. Since the experimental data produced have to be applicable to the full-scale prototype, the geometrical characteristics of the facility and the initial and boundary conditions of the selected tests have to be correctly scaled. Since possible scaling distortions are present in the experimental facility design, the similitude of the main thermal hydraulic phenomena of interest has to be assured permitting their accurate experimental simulation (Zuber, 1991; Reyes, 2005b; Reyes et al., 2007; Mascari et al., 2011e).

Different computer codes have been developed to characterize two-phase flow systems, from a system and a local point of view. Accurate simulation of transient system behavior of a nuclear power plant or of an experimental test facility is the goal of the best estimate thermal hydraulic system code. The evaluation of a thermal hydraulic system code’s calculation accuracy is accomplished by assessment and validation against appropriate system thermal hydraulic data, developed either from a running system prototype or from a scaled model test facility, and characterizing the thermal hydraulic phenomena during both steady state and transient conditions. The identification and characterization of the relevant thermal hydraulic phenomena, and the assessment and validation of thermal hydraulic systems codes, has been the objective of multiple international research programs (Mascari et al., 2011a; Mascari et al., 2011c).

In this international framework, Oregon State University (OSU) has constructed, under a U. S. Department of Energy grant, a system level test facility to examine natural circulation phenomena of importance to the MASLWR design. The scaling analysis of the OSU — MASLWR experimental facility was performed in order to have an adequately simulation of the single and two-phase natural circulation, reactor system depressurization during a blowdown and the containment pressure response typical of the MASLWR prototype (Zuber, 1991; Reyes & King, 2003; Reyes, 2005b). A previous testing program has been conducted in order to assess the operation of the prototypical MASLWR under normal full pressure and full temperature conditions and to assess the passive safety systems under transient conditions (Modro et al. 2003; Reyes & King, 2003; Reyes, 2005b; Reyes et al., 2007; Mascari et al., 2011e). The experimental data developed are useful also for the assessment and validation of the computational tools necessary for the operation, design and safety analysis of nuclear reactors.

For many years, in order to analyze the LWR reactors, the USNRC has maintained four thermal-hydraulic codes of similar, but not identical, capabilities, the RAMONA, RELAP5, TRAC-B and TRAC-P. In the last years, the USNRC is developing an advanced best estimate thermal hydraulic system code called TRAC/RELAP Advanced Computational Engine or TRACE, by merging the capabilities of these previous codes, into a single code (Boyac & Ward, 2000; TRACE V5.0, 2010; Reyes, 2005a; Mascari et al., 2011a). The validation and assessment of the TRACE code against the MASLWR natural circulation database, developed in the OSU-MASLWR test facility, is a novel effort.

This chapter illustrates an analysis of the primary/containment coupling phenomena characterizing the MASLWR design mitigation strategy during a SBLOCA scenario and, in the framework of the performance assessment and validation of thermal hydraulic system codes, a qualitative analysis of the TRACE V5 code capability in reproducing it.

Before commissioning and operation of the plant, a catalogue of thermal transients is compiled. These thermal transients are considered as design transients in contrast with the real transients based on temperature measured during operation. In the past, the anticipated transients were covering 40 years of plant operation. Now, the period to be covered is 60 years. Moreover, the specification is done for normal, upset, emergency and testing conditions. The design thermal transients are specified according to different plant models and experiences. They should always be conservative concerning frequency of occurrences, temperature range, rate of temperature change and load type (thermal stratification, thermal shock). Due to this conservatism, the usage factor calculated in the design phase, under normal circumstances, will be more severe than the results of the detailed fatigue calculation performed at a later operation stage taking into account the real operational thermal loads. As a consequence, usage factors around 1.0 are still tolerable in the design phase. They indicate the fatigue sensitive positions. These locations are selected for future instrumentation and non-destructive testing. Design improvements of components, depending on the calculated fatigue usage factors, can also be taken into account at this early stage. Additionally, optimization of operating modes can be considered. Thermal transients with low influence on fatigue behavior are identified as well. Depending on the different design codes [1], [4], [5], some procedures can allow the exemption of non significant loads. In the end, the predicted fatigue usage factors, which were calculated with design transients, shall be verified and the fatigue status shall be updated during lifetime operation.

The ultimate objective for advanced NPPs is to establish an enhanced approach to licensing, reflecting improved safety characteristics of advanced reactors, that is expected to justify and enable revised (reduced or eliminated) emergency planning requirements, while providing at least the same level to protection to the public as the current regulations. Ideally, the emergency planning zone would coincide with (or be contained within) the site boundary, thus, there would be no need for off-site evacuation planning, and the NPP would become, relative to the general population, the same type of facility as any other industrial enterprise.

In order to contribute toward achieving this ultimate objective by addressing some of the relevant issues there is a need to consider the following research tasks:

• Critically evaluate current regulations to identify what changes are necessary to enable advanced licensing.

• Identify criteria based on technical, quantifiable parameters that may be used in support of the objective.

• Identify approach, based on a combination of deterministic modelling, probabilistic analysis, and risk management, which will enable assessment of advanced plants based on their key design operational and safety characteristics with respect to adequate emergency planning requirements.

• Prepare site-specific representative data (e. g., meteorological).

• Perform probabilistic analyses needed to support the proposed approach.

• Perform deterministic / dose evaluation analyses needed to support the proposed approach.

• Perform a detailed evaluation of the representative reactor utilizing the combined proposed approach.

• Identify, discuss and quantify the benefits attainable through the implementation of this objective, i. e., licensing with reduced emergency planning requirements.

In order to perform these tasks with the ultimate goal of developing a technology — independent approach, the design of IRIS was used as a testbed. IRIS was representative of innovative reactors, but because it was a LWR, its possible sequences and its behaviour under accident conditions was much better understood and predicted than that of some more distant new technologies. Moreover, it had the necessary prerequisite, excellent safety, due to its Safety-by-DesignTM approach.

The related work was within the scope of activities defined within the International Atomic Energy Agency (IAEA) Co-ordinated Research Project (CRP) on Small Reactors with no or infrequent on-site refuelling. Specifically, it was relevant to "Definition of the scope of requirements and broader specifications" with respect to its ultimate objective (revised evacuation requirements) and to "Identification of requirements and broader specifications for NPPs for selected representative regions" considering specific impact on countries with colder climate and high interest for district heating co-generation.

It was expected that these results would contribute to ultimately defining a generic, country — independent approach and would support development of justification for reduced emergency planning through PRA analyses.

In addition, a study of the economic impact of revised licensing requirements on district heating was initiated. Thus the task was to perform economic study to evaluate positive economic effect on the nuclear district heating co-generation option, due to revised siting requirements with reduced emergency planning, which would allow placement of NPPs closer to population centres and allow them to be attractive option in heat energy supply market.

Finally, as part of this IAEA CRP a general methodology for revising the need for relocation and evacuation measures unique for NPPs for Innovative SMRs was developed and issued as IAEA publication (IAEA-TECDOC-1487, 2006).

Regarding further elaboration of the methodology it was suggested that external events and reasonable combinations of the external and internal events need to be included in the initial step of the methodology (accident sequence re-categorization), as for advanced reactors with the enforced inherent and safety by design features it might be that the impacts of external events would dominate the risk of severe accidents with possible radioactivity release. Work in this direction had already been started and was continued further, see (Alzbutas & Maioli, 2008) and (IAEA-TECDOC-1652, 2010).

(Price, 1995) remarks that Fung and Blevins have concluded that quasi-steady fluid

V

dynamics is valid provided fD -10 ; however, experiments by Price, Paidoussis and

Sychterz and others suggest that for closely spaced bodies the restriction on the use of quasi­steady fluid dynamics is much more severe than that suggested by Fung or Blevins. (Gross, 1975) carried out first quasi-steady analysis of cylinder arrays subjected to cross-flow concluding that instability in cylinder arrays is due to two distinct mechanisms: negative damping and stiffness controlled instability.

1.1.3 Computational fluid dynamic (CFD) models

The CFD solutions applicable to fluid-elastic instability and other problem areas of flow- induced vibrations are increasing. These include (Marn and Catton, 1991) and (Planchard & Thomas, 1993).

1.1.4 Non-linear models

The first non-linear model was given by (Roberts, 1962, 1966), who employed Krylov and Bogoliubov method (Minorsky, 1947) of averaging to solve the non-linear equations. Two — motivating forces have been remarked by (Price, 1995) for non-linear analyses. Firstly because of manufacturing tolerances and thermal constraints, there are likely to be small clearances between heat exchanger tubes and intermediate supports. Hence, large lengths of unsupported tubes, having very low natural frequencies. These low-frequencies may suffer from fluid-elastic instability at relatively low Vpc. A second and more academic motivating force for these non-linear analysis has been to investigate the possibility of Choatic behavior of tube motion.

1. The CSB assembly was aligned to the RV centerline and the CSB assembly was inserted in the RV. The CSB assembly was turned at 45° and was lowered to prevent damage to the DAK. It was then combined after the CSB assembly was turned to the original position before ensuring a 50 cm interval between the CSB assembly and the RV.

2. When the CSB assembly was installed, the load measured by a hydra-set was continuously checked. In addition, when the CSB assembly was at a height of approximately 30 cm from the RV head seating surface, the bottom surface (datum "B") of the CSB upper flange was used to stop the descent of the CSB assembly. A basis surface (datum "B") of the CSB upper flange was used for a parallel adjustment to within 0.381 mm of the RV head seating surface.

3. The load of the hydra-set was decreased to 10,000 lb and was checked given that the CSB came in contact with the RVI installation surface. The RV centerline and the CSB centerline were aligned within 0.0254 mm by CSB position devices (8 EA).

4. The vertical degree for the CSB keyway and the datum hole were measured and their relative positions on the CSB centerline were confirmed.

5. The gaps between the RV head seating surface and the upper surface of the CSB flange were measured in 45° intervals. The gaps (2.1336 — 2.9464 mm) of the RV outlet nozzle and the temperatures of the nozzle area were also measured.

6. The alignment of the RV / CSB centerline and the requirements of the nozzle gap were checked. If the requirements were not satisfied, it was necessary to repeat this procedure. If the position of the DAK changed before and after the installation of the CSB due to the checking of the position of the DAK, the measurements had to be done again and the existing checklist was invalidated. All installation requirements were met; the final adjustment conditions and the variation of the CSB centerline on the RV centerline were measured and recorded.

Rashydov Namik et al.[9]

Institute Cell Biology & Genetic Engineering of NAS of Ukraine Institute for Safety Problems of Nuclear Power Plants NAS of Ukraine National University "Kyievo-Mogiljanskaja Academy"

Ukraine

1. Introduction

The quarter century away pass after Chornobyl catastrophe. As result there surrounding lands object "Shelter" remain heavily contaminated by long-living radioactivity isotopes for many years to come. Nuclear danger dividing materials of object "Shelter" represent open sources of ionizing radiation (IR) and define not potential only, but direct danger to the personnel and environment. Today in object "Shelter" is about 95 % of highly active fuel loading of a reactor (an order 180 — 190 T of uranium and over 400 kg of plutonium). These danger materials are in different updating — in the form of active zone fragments, warm carried assemblages with the fulfilled nuclear fuel, lava-like fuel-containing materials (LFCM); in dispersion condition (a dust and aerosols); in water solutions of salts of uranium. Fuel containing materials represent congestions glassy state masses in the form of black, brown both polychromic ceramics and pumice state pieces of grey-brown colors. Virtually all materials of "Shelter" are sources of 90Sr, 137 Cs, 241Pu, appreciable amounts of plutonium isotopes (238,239,240Pu), as well as 241Am accumulated after the radoactivy decay of isotope 241Pu. Special attention found out and progressively developing intensive destruction of LFCM with formation highly dispersed "hot" particles (HP) on a surface.

It is shown, that dust generation ability of the fulfilled nuclear fuel of object "Shelter" from LFCM various type has high enough level both in the air environment, and in high vacuum. Annual total dust generation in object "Shelter" only at the expense of this mechanism it appears at level of several tens kg’s of the irradiated fuel. In HP activity averages P — and y- activity radionuclide decrease according to a half-life period about 30 years, and a-particles, on the contrary, increase at the expense of accumulation 241Am as disintegration product 241Pu additionally. HP from object "Shelter" are considered as the most radiation dangerous for biota because of the big factor of adjournment in breathe bodies (more than 40 %) and

the high radiating weighing factor (more than 20). Long track а-emitters in a living tissue (33 to 40 microns) significantly increases the radiation dosemore than 50 times due to appearance of local foci of exposure and increased risk of subsequent development of cancer as well as non-neoplastic diseases.

Dissolution of fuel containing materials in water inside "Shelter" as the result of ability to living microorganisms, results in occurrence of new bindings radionuclide — with organic substance are potentially dangereos and more mobile. On the basis of the received experimental data the biotic factor connection with radioactive aerosols in object "Shelter" is investigated. Existence of nano — and micro — size dust as radioactive aerosols with a number of specific properties inside the object "Shelter" and absence strong barriers against entered an environment of catching filters represents high potential risk of occurrence of adverse biomedical consequences for the personnel and for ecology of a 30-km Chornobyl zone.

The complex studies living mammalian organisms and of Chornobyl zone grown plants using post genomic methodologies such as genomics, transcriptomics and proteomics might provide detailed insight into the biochemistry of living plant cells influence chronic ionizing radiation were investigated. For plant of flax the main objective of this experimental research is to elucidate molecular compare changes between plants grown during flowering and embryogenesis in contaminated and control fields in Chornobyl area.

It is shown, that cells of different types (splenocytes, hepatocytes, bone marrow and astroglia cells) obtained from irradiated mice irrespective of a mode of influence of IR (total one-time external exposure y-radiation by 5.0 Sv, total external exposure y-radiation by 0.290 Sv for 231 days, long-term (over 74 days) incorporation of 137Cs with drink accumulation in the bodies of mice radioactivity (near 18 kBq) produce the factors not identified in this research in addition raising levels of single strand breaks (SSB) inside DNA for non-irradiated cells. In the conditions of a single exposition in у-fields with achievement of a dose of an external irradiation nearby 5.0 Sv intensity of production of «bystander» signals above at mice with the raised level of genetically determined sensitivity to response chronic irradiation. Under the same conditions of influence у-fields sheds light an induction of additional levels of SSB in DNA not irradiated cells on an extent at least one month after IR influence. The positive dynamics of serum levels of alanine aminotransferase and antibodies to the liver-specific lipoprotein — specific poly antigen for the liver, in the accumulation of dose 0.100 Sv is determined in animals that are under chronic exposure. Registered data correlated with pathomorphological changes in liver tissue.

An intra peritoneum injection of melanin with melanin-glucan complex from fungus Fomes fomentarius before irradiation procedure promotes essential decrease in production of «bystander» signals and normalization of hemopoietic progenitor cells, testifying in favor of free radical nature of their certain part. It is discovered also in our experimental research would be approved in Chornobyl zone the character of morphology changes develop, flowering, mature soybean seeds and flax of plant depend of chronic irradiation, changes in signal system and epigenetic changes as appear two peaks in curve depend of flowering rate during term of vegetation for flax which treatment with melanin content solution.

The researchers could use working models on base received experimental data in order to develop approach for living mammalian cells and plants which grown under influence chronic irradiation that were withstand consequences chronic irradiation of the radionuclide contamination environment around object "Shelter".

The steam generator with technical mark RGV-4E is one body SG [10-12]. The heat-exchange area is incorporated inside as surface of primary pipelines bundle with U-shape. The ends of these pipelines are fixed to the walls of the primary collector. Inside of SG body several separators and system of the steam water distribution are placed. The PGV-4E steam generator is foreseen for dry steam production with the pressure of about 4,61 MPa at a temperature of about 258°C.

The basic 1977 design from was improved after 1994 by new feed water pipeline system. There was also change in the type of steel of these pipelines. Instead conventional carbon steel, the austenite steel was used in distribution boxes as well as feed water pipelines.

All components in the Bohunice innovated feed water pipeline system were made of austenitic steel according to the Czechoslovak norm CSN, class 17. Advantages of the new construction are not only higher resistance against corrosion, but also much more comfortable visual inspection. The innovations can be seen in Figs. 1, 2.

The feed water comes via nozzle to distribution pipeline system and gets inside to left and right incoming line. From this place, water flows via pipelines ф 44,5x4mm into chambers and gets out via ejectors. This flow is mixed together with boiler water, so the final flow on the small primary pipelines is not extremely hot and does not cause a disturbing thermal load. Simultaneously, the circulation in SG tank was improved and places with increased salt concentration are reduced.

The main advantage is that the visual inspection of the feed water pipeline can be performed immediately due to placement of the whole water distribution system over the primary pipelines bundle. Using this system, the possible defects are easier observable. The next advantage is connected to 7 boxes with ejectors which mix properly the feed water with boiler water and the thermal load decreases in this way. An additional advantage is the checking and exchange possibility of distribution boxes in case of their damage.

Eugenijus Uspuras and Algirdas Kaliatka

Lithuanian Energy Institute Lithuania

1. Introduction

RBMK reactor belongs to the class of graphite-moderated nuclear power reactors that were designed in the Soviet Union in the 1950s. The usage of materials with low neutron absorption in RBMK design allows improving the fuel cycle by using cheap low-enriched nuclear fuel. In total 17 RBMK reactors have been built in Russia, Ukraine and Lithuania. One reactor is still under construction at Kursk Nuclear Power Plant (NPP). All three surviving reactors at Chernobyl NPP (Ukraine) were shutdown (the fourth was destroyed in the accident). Units 5 and 6 at Chernobyl NPP were under construction at the time of the accident; however, further construction was stopped due to the high contamination level at the site and political pressure. In Lithuania two reactors at Ignalina NPP were shutdown in 2004 and 2009. At present time no plans are made to build new RBMK type reactors, but in 2011, 11 RBMK reactors are still operating in Russia (4 reactors in Saint Petersburg, 3 — in Smolensk and 4 — in Kursk).

The RBMK reactor is a channel-type boiling water reactor. It has a huge graphite block structure, which functions as a moderator that slows down the neutrons produced by fission. The feature of RBMK type reactor is that each fuel assembly is positioned in its own vertical fuel channel, which is individually cooled by boiling water that is intended to remove the heat produced in it. The fuel channels are made of Zirconium and Niobium alloy similar to that used for fuel claddings. Reactor cooling system of RBMK has two loops, which are interconnected via the steamlines and do not have a connection on the water part. This is a difference from the vessel-type reactors.

The RBMK type reactors do not have full containment, preventing the environment from the radioactive material release. The absence of an overall containment suggests that in case of severe accident, the mitigation of fission products release to environment has to be based primarily on decreasing the extent of core damage, which is a key factor for the radiological consequences of accidents in RBMK. The degree of core damage is determined by the RBMK characteristics, such as the ability of the circulation loop to disintegrate and the multichannel nature of the core. Thus, depending on the type of accident, the damage of fuel assemblies can remain localized within a single fuel channel, a group of channels connected to the same group distribution header, or channels of a single loop (half of the core) or it can propagate to the entire core if complete loss of cooling occurs. Consequently, the severity of RBMK core damage depends on the degree and number of damaged fuel assemblies.

Another characteristic feature of RBMK is the graphite moderator. A positive property of such moderator is high heat capacity, which increases voided core heating time. This gives the operators more time to control the accident and to restore the failed equipment. At the same time, the existence of the graphite requires additional estimation of the graphite behavior at high temperature.

The mentioned specifics of RBMK reactors are affected on the design basis and beyond design basis accident sequences and necessary accident management measures, which are completely different from those in vessel type boiling water reactors. To understand the specifics of accidents in RBMK reactors the consequences of different accident groups were modeled by employing system thermal-hydraulic computer codes. This chapter presents the specifics of RBMK reactors, categorization of the Beyond Design Basis Accidents (BDBA) and specifics of the deterministic accident analyses in BDBA in RBMK. The results of the analysis were used for the development of Symptom-Based Emergency Operating Procedures and reactor cooldown strategies in case of beyond design basis accidents.