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Safety analysis of irradiated RBMK-1500 nuclear fuel transportation in newly developed container
Nuclear Engineering and Design 240 (2010) 3521–3528
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Nuclear Engineering and Design journal homepage: www.elsevier.com/locate/nucengdes
Safety analysis of irradiated RBMK-1500 nuclear fuel transportation in newly developed container S. Rimkeviˇcius ∗ , E. Uˇspuras, G. Dundulis, D. Laurinaviˇcius Lithuanian Energy Institute, Breslaujos 3, LT-44403 Kaunas, Lithuania
a r t i c l e
i n f o
Article history: Received 23 January 2010 Received in revised form 11 June 2010 Accepted 21 June 2010
a b s t r a c t Ignalina NPP (Lithuania) comprises two Units with RBMK-1500 reactors. After the Unit 1 of the Ignalina Nuclear Power Plant was shut down in 2004, approximately 1000 fuel assemblies from Unit 1 were available for further reuse in Unit 2. The fuel-transportation container, vehicle, protection shaft and other necessary equipment were designed in order to implement the process for on-site transportation of Unit 1 fuel for reuse in the Unit 2. The developed equipment can be used also in decommissioning phase for fuel transportation to fuel storage facilities. The set of this equipment can be applied for NPPs with RBMK type reactors. The purpose of this paper is to introduce the content and main results of safety analysis, focusing attention on the container used for spent fuel transportation. The structural integrity, thermal, radiological and nuclear criticality safety calculations are performed to assess the acceptance of the proposed set of equipment. The safety analysis demonstrated that the proposed nuclear fueltransportation container and other equipment are in compliance with functional, design and regulatory requirements, assure the required safety level for both normal operation and accident conditions and can be used for fuel transportation. © 2010 Elsevier B.V. All rights reserved.
1. Introduction RBMK is a water-cooled graphite-moderated channel-type power reactor. One of the specific features of the RBMK type reactor is that refueling is performed during power operation, the average burnup depth of fuel is kept constant during the operational life of the reactor, and after the reactor is shut down for decommissioning there are a substantial number of fuel assemblies with low burnup (considerable lower than design burnup depth). Such fuel assemblies have high energetic potential and can be reused. After Unit 1 of Ignalina was shut down in 2004, approximately 1000 fuel assemblies from Unit 1 were available for further reuse in Unit 2. Therefore, the project is implemented at Ignalina NPP in order to develop the process and set of equipment intended for reuse of Unit 1 fuel (which remains in the reactor after its shutdown) in the Unit 2 reactor. It is important to note that this approach has two benefits: it reduces the amount of fuel to be imported to Lithuania, and reduces the amount of radioactive waste to be stored. The safety assessment of equipment, intended for on-site transportation of RBMK fuel, is an issue of highest importance. Therefore, the safety analysis report is developed to demonstrate that the
∗ Corresponding author at: Lithuanian Energy Institute, Laboratory of Nuclear Installation Safety, Breslaujos 3, LT-44403 Kaunas, Lithuania. Tel.: +370 37 401924; fax: +370 37 351271. E-mail address: [email protected] (S. Rimkeviˇcius). 0029-5493/$ – see front matter © 2010 Elsevier B.V. All rights reserved. doi:10.1016/j.nucengdes.2010.06.027
proposed set of equipment performs all functions and assures the required level of safety. In the safety analysis report (SAR), the structural integrity, thermal, radiological and nuclear safety calculations are performed and the calculation results are compared with safety criteria and principles to assess the acceptance of the proposed set of equipment. The objective of this paper is to introduce the content and main results of safety analysis report, focusing attention on the safety justification for transportation container. The container is one of the most important components of the developed equipment set. Despite the fact that container is used for on-site transportation of RBMK fuel, the requirements for its design in some aspects are even more stringent than in the case of off-site transportation, because the fuel after the transportation shall be applicable for the reuse in Unit 2. The developed equipment can be used also in decommissioning phase for fuel transportation to fuel storage facilities. The set of this equipment can be applied for other NPPs with RBMK type reactors. 2. Description of process and equipment for Ignalina NPP Unit 1 fuel reuse in Unit 2 reactor In the first phase of the project, the process, which covers all the steps of transporting fuel from the Unit 1 reactor to the reactor of Unit 2, was developed. Technological process on fuel reuse can be divided into the following stages:
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Nomenclature keff GPS IAEA NPP RBMK SAR SFAs
effective neutron multiplication factor guide protective shaft International Atomic Energy Agency Nuclear Power Plant Russian acronym for ‘Large Channel Type WaterCooled Graphite-Moderated Reactor’ safety analysis report spent fuel assemblies
• preparation of Unit 1 spent fuel assemblies (SFAs) for transportation; • preparation of on-site transportation container for loading of the basket on Unit 1; • loading SFAs into the basket employing refueling machine on Unit 1; • loading filled basket into the transportation container on Unit 1; • preparation of on-site container for transportation on Unit 1; • transportation of on-site container from Units 1 to 2; • preparation of on-site transportation container for unloading/loading basket on Unit 2; • unloading loaded basket from container on Unit 2; • unloading SFAs from basket employing refueling machine on Unit 2; • loading basket into transportation container on Unit 2; • preparation of on-site container for transportation on Unit 2; • preparation of SFAs for loading in Unit 2; • SFAs loading into Unit 2. The fuel-transportation container, vehicle, protection shaft and other necessary equipment were designed to implement the proposed process. The guide protective shaft (GPS) represents the block construction composed of seven thick-walled cylindrical elements (Fig. 1)
Fig. 2. Transportation container on vehicle is ready for transporting fuel between Units at the Ignalina NPP site.
made of high-strength cast iron. Both Units 1 and 2 have a guide protective shaft. The GPS connects the transport corridor and reactor hall. The GPS is used to protect personnel from ionizing radiation and is the guiding path for transporting the basket from the container down to the level of reactor hall floor and back. The GPS has a block construction to achieve a uniform distribution of loadings to the building construction: each block of GPS has a separate support. The SFAs are transported from Units 1 to 2 in the container designed for the maintenance of radiation and nuclear safety at transportation of SFAs between power plant units. The container is placed on the transportation vehicle and is the steel thickwalled cylindrical vessel (Fig. 2), providing the arrangement of basket with 6 SFAs, protection from ionizing radiation, retention of radionuclides within the container and SFAs integrity during transportation. The transportation container is one of the most important components of the developed equipment set. Besides the requirements for compliance with required nuclear and radiation characteristics, the transportation container and all the associated equipment shall assure that the fuel will not be damaged during transportation and will be able to be reused in Unit 2. 3. Container safety assessment in the case of accidents 3.1. Content of safety analysis report
Fig. 1. Guide protective shaft.
The safety analysis report (SAR) begins with the detailed description of process and equipment, which is proposed and developed for on-site transportation of nuclear fuel at the Ignalina NPP. The safety classification of equipment in accordance with the Regulations of the Lithuanian Nuclear Safety Authority is provided in the following chapter of SAR. Then safety assessment of systems and equipment compliance with functional and design requirements as well as with safety standards and regulations (engineering assessment) is performed. A large part of the safety analysis report provides the assessment of container and other equipment operation at normal operational conditions and in the case of accidents. The structural integrity, thermal, radiological and nuclear safety calculations are performed, and the calculation results are compared with safety criteria and principles to assess the acceptance of the proposed set of equipment. Preparation of the justified list of the initial events (necessary for considering the substantiation of safety of the equipment providing transportation of fuel from Units 1 to 2) is carried out on the basis of the analysis of the suggested list of the initial events from Safety regulations PNAE G-14-029-91 (1991). All possible initiating events were considered and an in-depth analysis of these events was performed in order to justify the newly developed design of transportation container.
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The results of container safety assessment in the case of drop of fuel-loaded basket from the maximal possible height, as well as in the case of fire are presented in this paper in more detail. In addition to these accidents, the seismic event, shock wave, collision of the transportation vehicle and the container with an obstacle (railway vehicles or bays) along the line, loss of power supply, and human errors are considered in the safety analysis report and compliance with safety criteria is justified in case of all these events. 3.2. Structural integrity analysis for the case of drop of fuel-loaded basket from the maximum possible height During fuel assemblies loading (unloading) to (from) the transportation container the basket for fuel assemblies is fixed in the upper part of the protective shaft. The basket with fuel assemblies at this time is in 17 m distance above its position in container, and basket is not moved to any higher position during all technological process. So, the drop of a fuel-loaded basket from this maximal possible height of 17 m to the container is the most dangerous accident. This accident scenario is a design basis accident (DBA) regarding the developed set of equipment for fuel transportation. There is a number of preventive measures implemented in order to avoid such accident—e.g. the equipment used for transportation-technological operations with SFAs (cranes of reactor hall and spent fuel pools, gripping devices) exclude a possibility of spontaneous unhooking and the drop of SFAs and basket loaded with SFAs. However, the drop of a basket loaded with SFAs can be caused by the failure of these mechanisms as well as by human error. Therefore, a braking device was built into the flange of the basket to slow down the basket drop speed along the guide shaft. At the emergency drop of a basket loaded with fuel, the brake device operates automatically, which presses brake blocks to an internal surface of the guide protective shaft. The drop of basket into the container standing vertically is accompanied by the collapse of intra-container shock absorber, lowering the shock loading on the case of the container, the transportation vehicle, and the floor of the corridor. Thus the inertial loading affecting the SFAs essentially decreases. The strength calculations of the equipment in case of drop of fuel-loaded basket into container are performed in order to justify structural integrity of SFAs during this event. The allowable overload of fuel assemblies during the drop is determined by the strength of the carrier rod of each assembly, as the rods must remain intact in order to maintain structural integrity of SFAs. The strength calculations include a definition of resistance value and deformation of intra-container shock absorber, calculation of mechanical parameters behavior of basket starting from the moment of its loose, and including its movement through the shaft, dropping on the shock absorber, and up to its full stop in the container. One of most important components dealing with the consequences of drop of fuel-loaded basket is the shock absorber. The analysis of shock absorber performance includes both computational and experimental investigations, intended for justifying compliance with safety criteria (Karalevicius et al., 2006; Dundulis et al., 2009). The general view of the prepared shock absorber model is presented in Fig. 3. Static and dynamic study of the scaled shock absorber was performed. The purpose of the investigation of the scaled models for the shock absorber was verification of the construction of the shock absorber and subsequent validation of the developed finite-element model for calculation. Tests and simulations were carried out on scale models (1:5), geometrically similar to the full-scale shock absorber. The static analysis (Dundulis et al., 2009) was used for the evaluation of the design of shock absorber. The calculation methods were validated for the further use in the development of an optimum design of the shock absorber and for
Fig. 3. View of the scaled shock absorber model for testing.
the simulation of the behavior of the final design (non-scaled) of the shock absorber. The dynamic analysis was used for evaluation of the capability to withstand the dynamic load in case of an accident. The acceptability of the shock absorber was estimated according acceleration and displacement. The calculated acceleration (65.8 g) and displacement (61.8 mm) of the shock absorber are not exceeding the available acceleration and displacement (Karalevicius et al., 2006). It means that the absorber under design conditions is capable to absorb the energy from dropping the basket. Further, it is shown by calculations that the rod of fuel assembly does not break during drop of basket into container. Results of the carried out calculations show that all other constructive elements (spent fuel assembly, the container, transportation vehicle, protective shaft) also satisfy the strength conditions taking into account additional loading in case the basket is dropped in the container. The strength assessment of slab (floor) of the transport corridor in case of drop of basket filled with 6 SFAs was also performed. The loading affecting the slab of the transport corridor under normal operation will consist of the weight of all equipment (transportation vehicle, container, loaded basket), in total 1.53 MN (mega Newton). In an emergency situation, the dynamic component is added to this loading and the total loading on a floor of the transport corridor in case of drop of fuel-loaded basket into container is 2.14 MN. In the strength analysis of floor of the transport corridor it is conservatively assumed that loading on floor of the transport corridor without taking into account the operation of the brake device providing braking of basket will be 5 MN. For the analysis are used the standard mechanical characteristics of the concrete and reinforcement bar steel, and the characteristics of concrete obtained as a result of the investigations of the concrete condition in compartments of INPP Units 1 and 2, carried out within the framework of the project on Unit 1 fuel reuse in the reactor of Ignalina NPP Unit 2. The finite element methodology was used for strength analysis of reinforced concrete slabs. The finite-element model has been prepared for the analysis of slab of the transport corridor. The
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Fig. 5. Concrete failure indices dependence from loading.
Fig. 4. Finite-element model for the strength analysis of floor of the transport corridor.
finite-element model of building constructions has been developed using the finite element computer code ALGOR (Algor, 2000). Computer code ALGOR/NEPTUNE has been used for the transformation of all input data (nodes coordinates and materials properties) of ALGOR initial file into the initial file of code NEPTUNE (Dundulis et al., 1999). NEPTUNE is a three-dimensional finite element program developed to simulate the response of reactor components in three-dimensional space to design basis and beyond-design-basis loads. This computer code is specialized for the analysis of concrete structures. It was used reinforced concrete element for the analysis of reinforced concrete structures. Material nonlinearity such as concrete tension and compression, yielding of reinforcing steel are simulated with proper constitutive models. The code determines what layers are applied with tension and what are applied with compression and performs calculations up to the limit for tension and limit for compression. The computer code NEPTUNE (Kulak and Fiala, 1988) was used in order to handle the nonlinearities, such as concrete cracking and rebar plasticity, that occur during this extreme loading event. The slabs and walls of the building were modeled using an enhanced version of the four-node quadrilateral plate element developed by Belytschko et al. (1984) for metallic structures. The formulation of this element is based upon the Mindlin theory of plates and uses a velocity strain formulation. The element was further developed by Kulak and Fiala (1989) and Kulak et al. (1997) by incorporating the features to represent concrete and reinforcing steel. Subsequently, additional failure criteria were added, and this enabled the modified element to model concrete cracking, reinforcing bar failure and gross transverse failure. The concrete failure model used is the Hsieh-Ting-Chen four-parameter model, so that to represent the following four failure states (Kulak et al., 1997): 1. 2. 3. 4.
this case the loading from weight of the vehicle with the container, and also loading from drop of basket filled with 6 SFAs is imposed on points of contact of transportation vehicle wheels with rails (Fig. 4, loading is presented by arrows), and no loading is carried by jacks. The calculations performed for the standard mechanical characteristics of slab showed that the destructions of floor will not occur, but formation of cracks in concrete will start at loading of floor of the transport corridor. Fig. 5 shows the loading evolution of the concrete failure index (CFAIL) for each layer in most loaded element. The tensile failure surface of concrete is reached at layers 3–5, which are located in the tension side of the floor, and this is shown in Fig. 5 by CFAIL = 1 (limit for tension is reached in Layer 3, where the concrete cracking begins, and only tensile failure is reached) and CFAIL = 2 (tensile and compressive failure surface is reached—tension evaluation is completed and the crack in concrete starts to open). Layers 1–2 have CFAIL = 0 (no failure), and this indicates that neither the tensile nor compression failure surface has not been reached. The failure of the concrete begin at 400 t load at layers 3–5. The experimental (testing) mechanical characteristics of concrete obtained as a result of special investigations is much better than standard. The NEPTUNE analysis results using experimental properties of concrete and reinforcing bars material are presented in Fig. 6. The maximum normal stress in concrete is 1.37 MPa (see Fig. 6). It means that in concrete of floor ultimate strength to tension is not achieved and opening of cracks does not begin (tension, tensile strength is 3.7 MPa). The calculations performed for experimental mechanical characteristics of concrete prove that the destructions of slab will not occur, also the formation of cracks in concrete will not start at loading of floor of the transport corridor by weight of 500 tons (5 MN), i.e., floor maintains this loading.
uniaxial compressive strength, uniaxial tensile strength, equal biaxial compressive strength, combined triaxial compression.
NEPTUNE calculates stresses on central axis of element in five integration points for concrete and in each rebar layer. In order to evaluate structural integrity of slab during drop of fuel-loaded basket from a height of 17 m, conservatively an additional failure of jacks of the transportation vehicle was assumed. In
Fig. 6. Maximum normal stress (Pa) distribution in concrete of the slab.
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The strength analysis of floor of the transport corridor showed that the slab will sustain drop of fuel-loaded basket in the case with the failure of jacks. 3.3. Safety justification of container in the case of fire The performed calculations demonstrated that decay heat is efficiently removed from SFAs to the container vessel and further to the environment by the natural convection and radiation and during fuel transportation at normal operation conditions the temperature of fuel roads does not exceed 153 ◦ C. To meet the requirements of regulations, fuel roads shall not exceed the permissible temperature (300 ◦ C) not only for normal operation conditions, but also during the fire. Therefore, the safety analysis of fuel transportation in the case of fire is also provided in the safety analysis report. For thermal characteristics calculation in the case of fire, the parameters of fire are accepted in accordance with the Regulations OPBZ-83 (1983) and TS-R-1 (2009) when fire temperature is equal to 800 ◦ C, whereas duration of fire is 0.5 h. Such fire could be caused by the spilt of liquid fuel during the accident. In addition, as the fire parameters in case of container storage, flame temperature is assumed to be 600 ◦ C, whereas fire duration—1 h, like in the case of the simulation of the CONSTOR RBMK-1500 container in the case of fire. The calculation of thermal characteristics of the container is carried out using the method of finite elements employing three-dimensional programme ANSYS version 5.5.1. Designed finite-element model of the container contains 3789 finite elements and 3545 junctions. Calculation results show the available margin to the maximum permissible temperature of fuel rods in the case of fire. In case of container appearance in the fireplace with the flame temperature of 600 ◦ C, at fire duration of 1 h and with the subsequent cooling in the air, the maximum temperature of fuel rods is achieved in 4.3 h after entering the fireplace and comprises 188 ◦ C. In case the container enters the fireplace with temperature of 800 ◦ C at fire duration of 0.5 h and with the subsequent cooling in the air, the maximum temperature of fuel rods is achieved in 3.7 h after entering the fireplace and comprises 201 ◦ C. Level of temperatures of metal elements in the case of fire (approximately 200 ◦ C) and insignificant duration of such temperatures do not have a negative impact on the equipment in general. The simulation showed that maximum temperature gradients occurring in thick-walled elements of the casing in especially unfavorable time spans and due to that obtained thermal stresses do not reduce the strength of these constructive elements. Obtained calculation results confirmed that in case of transportation container entrance into fire the failure of fuel cladding integrity does not occur. 3.4. Nuclear criticality safety The original design calculations for nuclear criticality safety evaluation for spent nuclear fuel transport container were performed by employing codes MCNP and (KENO-VI). In this paper the results of verification calculations are presented and compared with design calculations. Verification calculations were performed using SCALE (USA) computer code, which uses the criticality program KENO-V (Petrie et al., 2006). In the verification calculations nuclear criticality analysis of spent nuclear fuel container was made in several stages. The basket with six fuel assemblies is located in the container. First of all the model of RBMK-1500 fuel assembly was developed (Fig. 7) and then it was included in the model of container with the basket for 6 SFAs. The RBMK-1500 fuel assembly consists of two segments (upper and lower fuel bundles), 18 fuel each. Fuel elements are arranged in two concentric rings with a central carried rod (Fig. 7).
Fig. 7. RBMK-1500 fuel assembly digital model.
The geometrical characteristics of fuel assembly are presented in the Ignalina NPP Source book (Almenas et al., 1998) and are used in the model developed by employing SCALE computer code. Spent nuclear fuel container criticality analysis was made in several stages by employing the SCALE code. First, the geometrical model of the RBMK-1500 fuel assembly was made. Then was modeled fresh fuel burning process, using SAS2H module of SCALE code. During fuel burning process a new unstable nuclides are generated. In the second stage, the obtained new fuel nuclides composition was used for the criticality calculation. The study aims to verify whether the criticality value does not exceed the allowable limits based on IAEA documents. The same methodology was used by Calic and Ravnik (2010). They carried out calculations with MCNP code. At this work were investigated 3 different configurations of spent nuclear fuel assemblies in baskets of the container (Fig. 8). First case represents normal transportation configuration, and the second with the third are hypothetical conservative cases then the spent fuel assemblies are sagged to critical mass. The criticality calculations for all three configurations were made at the different water concentrations (0.01–1.00 g/cm3 ) and fuel enrichments (2.0% and 2.8% U235 ). As shown in Fig. 9 the highest criticality values are received, then the fuel assemblies are sagged to one place (see 3rd case in Fig. 8) and the fuel enrichment is 2.8%. In this case criticality value could reach about 0.7 in the case of water density 1.00 g/cm3 . However in reality not more than 10 l of water can appear in the container during fuel loading process. This corresponds to the average water density of 0.14 g/cm3 and criticality value in this case does not exceed 0.4. The received nuclear criticality safety results were compared with the results of design calculations. Fig. 10 presents the comparison of criticality calculation results received by different computer codes and different authors for normal transportation configuration in the case of 2.8% fuel enrichment (i.e., maximal enrichment for RBMK-1500 fuel). As we can see in Fig. 10 the results of verification calculations performed by the SCALE (KENO-V) code are very similar to the results of design calculations received by KENOVI and MCNP computer codes. The difference between different results is just about 2%. Thus, the performed calculations proved that a nuclear criticality safety is assured with large margin, because the acceptance criterion keff = 0.95 is not achieved for both normal operation and accident conditions. 3.5. Radiological safety 3.5.1. Radiological safety during normal operational conditions The original design calculations of sum dose rates for spent nuclear fuel transport container were performed by employing code MCNP. In this paper the results of verification calculations
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Fig. 9. The criticality dependence from water density of the container with 6 RBMK1500 nuclear fuel assemblies (1st case, 2nd case and 3rd case are the arrangements of spent nuclear fuel assemblies in the container—see Fig. 8).
are presented and compared with design calculations. Verification calculations were performed using SCALE (USA) computer code, which uses the sum dose rate calculation module SAS 4 (Tang and Emmett Margaret, 2006). The investigation was made for the different spent nuclear fuel cooling times (1, 1.5, 3, 5 years) in the water pools, and different burnup levels (4.5, 5, 10, 15, 15.6, and 18.9 MWd/kgU). The sum dose rates were calculated on the external container surface and at different distances from this surface. The calculated sum dose includes gamma and neutrons doses. As already mentioned in Section 3.4, first of all fresh fuel burning process was modeled. New unstable nuclides are generated during the fuel burning process, and a certain amount of energy is emitted (energy depend from the fuel burnup level). The obtained amount of energy was used for dose rate calculations of spent nuclear fuel container. For dose rate calculations spent fuel container was simulated by 3 zones (Fig. 11). The central inner part of geometrical model of spent fuel container represents six RBMK-1500 fuel assemblies, which were homogenized in one zone (Zone 1 in Fig. 11). Other zone of model represents single area of air space (Zone 2 in Fig. 11) and a container wall (Zone 3). The same methodology was used by Margeanu and Angelescu (2003). The homogenous model of container with nuclear fuel was developed for dose rate assessment. Six fuel assemblies were homogenized to one zone in this model (Fig. 11).
Fig. 8. The arrangements of spent nuclear fuel assemblies in transport container baskets. Fig. 10. The transport container with 2.8% U235 fuel enrichment criticality dependence from water density (1st case the arrangements of spent nuclear fuel assemblies in the container—see Fig. 8).
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Fig. 13. Comparison of verification calculations (SCALE) and design results (MCNP). Fig. 11. Model of transport container with spent fuel: 1, homogenous fuel zone; 2, void; 3, container wall (steel).
Fig. 14. Calculation results comparison with measurements. Fig. 12. Dependence of sum dose rate on the external container surface from the spent burnup level and fuel cooling time.
The results of dose rate on the external container surface show that dose rate is largest when the cooling time is the smallest and burnup level is the highest (Fig. 12). The SCALE calculation results of dose rates on the external container surface and at different distances from this surface (0.5, 1.0, and 2.0 m) were compared with results of design calculations performed by the MCNP computer code. In accordance with boundary conditions and results of optimization calculations the spent nuclear fuel is transported by the container after at least 1.5 years cooling in the water pool, whose maximal burnup level does not exceed 18.9 MWd/kgU and initial enrichment of U235 is 2.6%. Fig. 13 presents results received for these most conservative conditions regarding nuclear fuel cooling time and burnup level. The comparison showed that calculated dose rates are very similar, the difference between SCALE and MCNP results does no exceed 5% (Fig. 13). Fig. 13 also shows that the maximal calculated dose rate value on the container external surface is 214 Sv/h for the most conservative conditions regarding nuclear fuel cooling time, enrichment and burnup level. This calculated value is almost ten times lower even than the limit of 2 mSv/h, which is defined in the International Atomic Energy Agency Safety Standard TS-R-1 (2009) as the maximum radiation level at any point on the external surface of a package for radioactive material transportation on a public roads and railways (in our case the container is used for on-site
fuel transportation). The radiation levels at a distances of 1 m and 2 m from the external surface of the package is also well below the limits prescribed by the IAEA Safety requirements No. TS-R-1 (2009). The calculation results were compared also with data of experimental measurements of dose rates on the external container surface and at the different distances from it for the fuel burnup levels of 4.5 and 15.6 MWd/kgU. Fig. 14 shows that calculation values are at ∼30% higher than measurement values. Thus, it can be concluded that calculated values are conservative and assures additional safety margin. Broadhead et al. (1997) obtained about 30% relative error of measurement comparing with calculation values. Ene (2000) pointed important errors (up to 30%) caused by modeling limitations. Thus, the other authors received similar differences between measurements and calculations as presented in this paper. 3.5.2. Radiological safety in the case of beyond design basis accident Strength analysis calculations confirmed that there is no fuel rods damage in case of drop of full-loaded basket (i.e., with 6 SFAs) from the maximum possible height (17 m), even in the case of single failure of one of the elements of basket brake device (i.e., in the case of design basis accident with single failure). For the assessment of radiological consequences the beyond design basis accident scenario – drop of full-loaded basket from the maximum possible height with simultaneous complete failure of basket brake device (simultaneous failure of all brake blocks) – was analyzed. It was
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assumed that in this scenario the loss of leak tightness of all fuel rods in all 6 SFAs occurs. Thus release of radioactive nuclides located in gas spaces under fuel rod claddings will occur. The carried out assessment of radiological consequences on the border of INPP controlled area (3 km) has shown that as a result of analyzed accident scenario the radiation dose for the population (taking into account all ways of an irradiation) for the first year after the accident will not exceed 11 Sv, that is significantly below than the annual effective limited dose of the population for operation conditions—200 Sv (HN 87, 2002) and, thus, intervention on the protection of the population is not required in accordance with regulations (HN 73, 2001). 4. Conclusions The safety analysis report demonstrated that the proposed transportation container and other equipment, which are intended for on-site transportation of RBMK reactor fuel, are in compliance with functional, design and regulation requirements and assure the required safety level for both normal operation and accident conditions. The calculation demonstrated that maximum permissible temperature of fuel rods (300 ◦ C) is not achieved even in the case of fire (maximal temperature is ∼200 ◦ C). Results of the calculations show that all constructive elements (spent fuel assembly, the container) maintain their structural integrity even in the case of dropping a fuel-loaded basket from the maximal possible height (17 m). Radiological and nuclear safety criteria are met for both normal operation and accident conditions. The developed equipment can be used also in decommissioning phase for fuel transportation to fuel storage facilities. The set of this equipment can be applied for other NPPs with RBMK type reactors. Acknowledgments The authors want to extend thanks to the administration and technical staff at the Ignalina NPP, for providing information regarding operational procedures and operational data.
References Algor, 2000. Algor Finite Element Analysis System. ALGOR Instruction Manuals, Inc., Pittsburgh. Almenas, K., Kaliatka, A., Uspuras, E., 1998. Ignalina RBMK-1500, A Source Book. Lithuanian Energy Institute, Kaunas. Belytschko, T., Lin, J.I., Tsay, C.S., 1984. J. Comput. Methods Appl. Mech. Eng. 42, 225–251. Broadhead, B.L, Tang, J.S., Childs, R.L., Parks, C.V., Taniuchi, H., 1997. Evaluation of shielding analysis methods in spent-fuel cask environments. Nucl. Technol. 117 (2), 206–222. Calic, D., Ravnik, M., 2010. Criticality calculation of spent fuel in deep geological repository. Nucl. Eng. Des. 240 (3), 668–671. Dundulis, G., Narvydas, E., Uspuras, E., Listopadskis, N., 1999. Confinement study using Algor and Neptune codes. In: Transactions of 15th International Conference on Structural Mechanics in Reactor Technology, Seoul, Korea, VI, p. 333. ˙ Dundulis, G., Grybenas, A., Karaleviˇcius, R., Makareviˇcius, V., Rimkeviˇcius, S., Uˇspuras, E., 2009. Static analytical and experimental research of shock absorber to safeguard the nuclear fuel assemblies. Nucl. Eng. Des. 239 (1), 1–8. Ene, D., 2000. Dose rate estimates exterior to the castor MTR-2 cask used for storage of the WWR-S Romanian reactor spent fuel. In: Safety of Radioactive Waste Management, Vienna IAEA-CN-78/32, pp. 1–5. HN 73, 2001. Hygiene Standard. Basic Standard of Radiation Protection, Vilnius. HN 87, 2002. Hygiene Standard. Radiation Protection in Nuclear Facilities, Vilnius. Karalevicius, R., Dundulis, G., Rimkevicius, S., Uspuras, E., 2006. Investigation of a shock absorber for safeguard of fuel assemblies failure. In: Proc. of 14th International Conference on Nuclear Engineering (ICONE14), Miami, USA, July 17–20, 2005, CD-ROM, 8 p. Kulak, R.F., Pfeiffer, P.A., Plaskacz, E.J., 1997. Modeling of containment structures on high performance computers. Nucl. Eng. Des. 174 (2), 143–156. Kulak, R.F., Fiala, C., 1988. NEPTUNE: a system of finite element programs for three dimensional nonlinear analysis. Nucl. Eng. Des. 106, 47–68. Kulak, R.F., Fiala, C., 1989. A method for three-dimensional structural analysis of reinforced concrete containment. In: Shin, Y.S., Wang, C.Y., Colton, J.D., Kulak, R.F. (Eds.), Shock and Wave Propagation, Fluid-Structure Interaction and Structural Response, vol. 159, ASME PVP Conf. July 1989, pp. 115–123. Margeanu, C.A., Angelescu, T., 2003. Shielding calculations for spent CANDU fuel transport cask. Rom. Rep. Phys. 55 (3), 219–223. OPBZ-83, 1983. Basic regulations for safety and physical protection of nuclear materials transportation. Petrie, L.M., Landers, N.F., Hollenbach, D.F., 2006. Rearden B.T., KENO V.a: An Improved Monte Carlo Criticality Program, Rev. 5.1, ORNL/TM-2005/39. PNAE G-14-029-91, 1991. Safety regulations at storage and transportation of nuclear fuel in nuclear energy objects. Tang, J.S., Emmett Margaret, B., 2006. SAS4: A Monte Carlo cask shielding analysis module using an automated biasing procedure, Ver. 5.1 ORNL/TM-2005/39. TS-R-1, 2009. Regulations for the safe transport of radioactive material. In: Safety Requirements, 2009 Edition, IAEA Safety Standards Series. International Atomic Energy Agency, Vienna.
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Nuclear Engineering and Design journal homepage: www.elsevier.com/locate/nucengdes
Safety analysis of irradiated RBMK-1500 nuclear fuel transportation in newly developed container S. Rimkeviˇcius ∗ , E. Uˇspuras, G. Dundulis, D. Laurinaviˇcius Lithuanian Energy Institute, Breslaujos 3, LT-44403 Kaunas, Lithuania
a r t i c l e
i n f o
Article history: Received 23 January 2010 Received in revised form 11 June 2010 Accepted 21 June 2010
a b s t r a c t Ignalina NPP (Lithuania) comprises two Units with RBMK-1500 reactors. After the Unit 1 of the Ignalina Nuclear Power Plant was shut down in 2004, approximately 1000 fuel assemblies from Unit 1 were available for further reuse in Unit 2. The fuel-transportation container, vehicle, protection shaft and other necessary equipment were designed in order to implement the process for on-site transportation of Unit 1 fuel for reuse in the Unit 2. The developed equipment can be used also in decommissioning phase for fuel transportation to fuel storage facilities. The set of this equipment can be applied for NPPs with RBMK type reactors. The purpose of this paper is to introduce the content and main results of safety analysis, focusing attention on the container used for spent fuel transportation. The structural integrity, thermal, radiological and nuclear criticality safety calculations are performed to assess the acceptance of the proposed set of equipment. The safety analysis demonstrated that the proposed nuclear fueltransportation container and other equipment are in compliance with functional, design and regulatory requirements, assure the required safety level for both normal operation and accident conditions and can be used for fuel transportation. © 2010 Elsevier B.V. All rights reserved.
1. Introduction RBMK is a water-cooled graphite-moderated channel-type power reactor. One of the specific features of the RBMK type reactor is that refueling is performed during power operation, the average burnup depth of fuel is kept constant during the operational life of the reactor, and after the reactor is shut down for decommissioning there are a substantial number of fuel assemblies with low burnup (considerable lower than design burnup depth). Such fuel assemblies have high energetic potential and can be reused. After Unit 1 of Ignalina was shut down in 2004, approximately 1000 fuel assemblies from Unit 1 were available for further reuse in Unit 2. Therefore, the project is implemented at Ignalina NPP in order to develop the process and set of equipment intended for reuse of Unit 1 fuel (which remains in the reactor after its shutdown) in the Unit 2 reactor. It is important to note that this approach has two benefits: it reduces the amount of fuel to be imported to Lithuania, and reduces the amount of radioactive waste to be stored. The safety assessment of equipment, intended for on-site transportation of RBMK fuel, is an issue of highest importance. Therefore, the safety analysis report is developed to demonstrate that the
∗ Corresponding author at: Lithuanian Energy Institute, Laboratory of Nuclear Installation Safety, Breslaujos 3, LT-44403 Kaunas, Lithuania. Tel.: +370 37 401924; fax: +370 37 351271. E-mail address: [email protected] (S. Rimkeviˇcius). 0029-5493/$ – see front matter © 2010 Elsevier B.V. All rights reserved. doi:10.1016/j.nucengdes.2010.06.027
proposed set of equipment performs all functions and assures the required level of safety. In the safety analysis report (SAR), the structural integrity, thermal, radiological and nuclear safety calculations are performed and the calculation results are compared with safety criteria and principles to assess the acceptance of the proposed set of equipment. The objective of this paper is to introduce the content and main results of safety analysis report, focusing attention on the safety justification for transportation container. The container is one of the most important components of the developed equipment set. Despite the fact that container is used for on-site transportation of RBMK fuel, the requirements for its design in some aspects are even more stringent than in the case of off-site transportation, because the fuel after the transportation shall be applicable for the reuse in Unit 2. The developed equipment can be used also in decommissioning phase for fuel transportation to fuel storage facilities. The set of this equipment can be applied for other NPPs with RBMK type reactors. 2. Description of process and equipment for Ignalina NPP Unit 1 fuel reuse in Unit 2 reactor In the first phase of the project, the process, which covers all the steps of transporting fuel from the Unit 1 reactor to the reactor of Unit 2, was developed. Technological process on fuel reuse can be divided into the following stages:
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Nomenclature keff GPS IAEA NPP RBMK SAR SFAs
effective neutron multiplication factor guide protective shaft International Atomic Energy Agency Nuclear Power Plant Russian acronym for ‘Large Channel Type WaterCooled Graphite-Moderated Reactor’ safety analysis report spent fuel assemblies
• preparation of Unit 1 spent fuel assemblies (SFAs) for transportation; • preparation of on-site transportation container for loading of the basket on Unit 1; • loading SFAs into the basket employing refueling machine on Unit 1; • loading filled basket into the transportation container on Unit 1; • preparation of on-site container for transportation on Unit 1; • transportation of on-site container from Units 1 to 2; • preparation of on-site transportation container for unloading/loading basket on Unit 2; • unloading loaded basket from container on Unit 2; • unloading SFAs from basket employing refueling machine on Unit 2; • loading basket into transportation container on Unit 2; • preparation of on-site container for transportation on Unit 2; • preparation of SFAs for loading in Unit 2; • SFAs loading into Unit 2. The fuel-transportation container, vehicle, protection shaft and other necessary equipment were designed to implement the proposed process. The guide protective shaft (GPS) represents the block construction composed of seven thick-walled cylindrical elements (Fig. 1)
Fig. 2. Transportation container on vehicle is ready for transporting fuel between Units at the Ignalina NPP site.
made of high-strength cast iron. Both Units 1 and 2 have a guide protective shaft. The GPS connects the transport corridor and reactor hall. The GPS is used to protect personnel from ionizing radiation and is the guiding path for transporting the basket from the container down to the level of reactor hall floor and back. The GPS has a block construction to achieve a uniform distribution of loadings to the building construction: each block of GPS has a separate support. The SFAs are transported from Units 1 to 2 in the container designed for the maintenance of radiation and nuclear safety at transportation of SFAs between power plant units. The container is placed on the transportation vehicle and is the steel thickwalled cylindrical vessel (Fig. 2), providing the arrangement of basket with 6 SFAs, protection from ionizing radiation, retention of radionuclides within the container and SFAs integrity during transportation. The transportation container is one of the most important components of the developed equipment set. Besides the requirements for compliance with required nuclear and radiation characteristics, the transportation container and all the associated equipment shall assure that the fuel will not be damaged during transportation and will be able to be reused in Unit 2. 3. Container safety assessment in the case of accidents 3.1. Content of safety analysis report
Fig. 1. Guide protective shaft.
The safety analysis report (SAR) begins with the detailed description of process and equipment, which is proposed and developed for on-site transportation of nuclear fuel at the Ignalina NPP. The safety classification of equipment in accordance with the Regulations of the Lithuanian Nuclear Safety Authority is provided in the following chapter of SAR. Then safety assessment of systems and equipment compliance with functional and design requirements as well as with safety standards and regulations (engineering assessment) is performed. A large part of the safety analysis report provides the assessment of container and other equipment operation at normal operational conditions and in the case of accidents. The structural integrity, thermal, radiological and nuclear safety calculations are performed, and the calculation results are compared with safety criteria and principles to assess the acceptance of the proposed set of equipment. Preparation of the justified list of the initial events (necessary for considering the substantiation of safety of the equipment providing transportation of fuel from Units 1 to 2) is carried out on the basis of the analysis of the suggested list of the initial events from Safety regulations PNAE G-14-029-91 (1991). All possible initiating events were considered and an in-depth analysis of these events was performed in order to justify the newly developed design of transportation container.
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The results of container safety assessment in the case of drop of fuel-loaded basket from the maximal possible height, as well as in the case of fire are presented in this paper in more detail. In addition to these accidents, the seismic event, shock wave, collision of the transportation vehicle and the container with an obstacle (railway vehicles or bays) along the line, loss of power supply, and human errors are considered in the safety analysis report and compliance with safety criteria is justified in case of all these events. 3.2. Structural integrity analysis for the case of drop of fuel-loaded basket from the maximum possible height During fuel assemblies loading (unloading) to (from) the transportation container the basket for fuel assemblies is fixed in the upper part of the protective shaft. The basket with fuel assemblies at this time is in 17 m distance above its position in container, and basket is not moved to any higher position during all technological process. So, the drop of a fuel-loaded basket from this maximal possible height of 17 m to the container is the most dangerous accident. This accident scenario is a design basis accident (DBA) regarding the developed set of equipment for fuel transportation. There is a number of preventive measures implemented in order to avoid such accident—e.g. the equipment used for transportation-technological operations with SFAs (cranes of reactor hall and spent fuel pools, gripping devices) exclude a possibility of spontaneous unhooking and the drop of SFAs and basket loaded with SFAs. However, the drop of a basket loaded with SFAs can be caused by the failure of these mechanisms as well as by human error. Therefore, a braking device was built into the flange of the basket to slow down the basket drop speed along the guide shaft. At the emergency drop of a basket loaded with fuel, the brake device operates automatically, which presses brake blocks to an internal surface of the guide protective shaft. The drop of basket into the container standing vertically is accompanied by the collapse of intra-container shock absorber, lowering the shock loading on the case of the container, the transportation vehicle, and the floor of the corridor. Thus the inertial loading affecting the SFAs essentially decreases. The strength calculations of the equipment in case of drop of fuel-loaded basket into container are performed in order to justify structural integrity of SFAs during this event. The allowable overload of fuel assemblies during the drop is determined by the strength of the carrier rod of each assembly, as the rods must remain intact in order to maintain structural integrity of SFAs. The strength calculations include a definition of resistance value and deformation of intra-container shock absorber, calculation of mechanical parameters behavior of basket starting from the moment of its loose, and including its movement through the shaft, dropping on the shock absorber, and up to its full stop in the container. One of most important components dealing with the consequences of drop of fuel-loaded basket is the shock absorber. The analysis of shock absorber performance includes both computational and experimental investigations, intended for justifying compliance with safety criteria (Karalevicius et al., 2006; Dundulis et al., 2009). The general view of the prepared shock absorber model is presented in Fig. 3. Static and dynamic study of the scaled shock absorber was performed. The purpose of the investigation of the scaled models for the shock absorber was verification of the construction of the shock absorber and subsequent validation of the developed finite-element model for calculation. Tests and simulations were carried out on scale models (1:5), geometrically similar to the full-scale shock absorber. The static analysis (Dundulis et al., 2009) was used for the evaluation of the design of shock absorber. The calculation methods were validated for the further use in the development of an optimum design of the shock absorber and for
Fig. 3. View of the scaled shock absorber model for testing.
the simulation of the behavior of the final design (non-scaled) of the shock absorber. The dynamic analysis was used for evaluation of the capability to withstand the dynamic load in case of an accident. The acceptability of the shock absorber was estimated according acceleration and displacement. The calculated acceleration (65.8 g) and displacement (61.8 mm) of the shock absorber are not exceeding the available acceleration and displacement (Karalevicius et al., 2006). It means that the absorber under design conditions is capable to absorb the energy from dropping the basket. Further, it is shown by calculations that the rod of fuel assembly does not break during drop of basket into container. Results of the carried out calculations show that all other constructive elements (spent fuel assembly, the container, transportation vehicle, protective shaft) also satisfy the strength conditions taking into account additional loading in case the basket is dropped in the container. The strength assessment of slab (floor) of the transport corridor in case of drop of basket filled with 6 SFAs was also performed. The loading affecting the slab of the transport corridor under normal operation will consist of the weight of all equipment (transportation vehicle, container, loaded basket), in total 1.53 MN (mega Newton). In an emergency situation, the dynamic component is added to this loading and the total loading on a floor of the transport corridor in case of drop of fuel-loaded basket into container is 2.14 MN. In the strength analysis of floor of the transport corridor it is conservatively assumed that loading on floor of the transport corridor without taking into account the operation of the brake device providing braking of basket will be 5 MN. For the analysis are used the standard mechanical characteristics of the concrete and reinforcement bar steel, and the characteristics of concrete obtained as a result of the investigations of the concrete condition in compartments of INPP Units 1 and 2, carried out within the framework of the project on Unit 1 fuel reuse in the reactor of Ignalina NPP Unit 2. The finite element methodology was used for strength analysis of reinforced concrete slabs. The finite-element model has been prepared for the analysis of slab of the transport corridor. The
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Fig. 5. Concrete failure indices dependence from loading.
Fig. 4. Finite-element model for the strength analysis of floor of the transport corridor.
finite-element model of building constructions has been developed using the finite element computer code ALGOR (Algor, 2000). Computer code ALGOR/NEPTUNE has been used for the transformation of all input data (nodes coordinates and materials properties) of ALGOR initial file into the initial file of code NEPTUNE (Dundulis et al., 1999). NEPTUNE is a three-dimensional finite element program developed to simulate the response of reactor components in three-dimensional space to design basis and beyond-design-basis loads. This computer code is specialized for the analysis of concrete structures. It was used reinforced concrete element for the analysis of reinforced concrete structures. Material nonlinearity such as concrete tension and compression, yielding of reinforcing steel are simulated with proper constitutive models. The code determines what layers are applied with tension and what are applied with compression and performs calculations up to the limit for tension and limit for compression. The computer code NEPTUNE (Kulak and Fiala, 1988) was used in order to handle the nonlinearities, such as concrete cracking and rebar plasticity, that occur during this extreme loading event. The slabs and walls of the building were modeled using an enhanced version of the four-node quadrilateral plate element developed by Belytschko et al. (1984) for metallic structures. The formulation of this element is based upon the Mindlin theory of plates and uses a velocity strain formulation. The element was further developed by Kulak and Fiala (1989) and Kulak et al. (1997) by incorporating the features to represent concrete and reinforcing steel. Subsequently, additional failure criteria were added, and this enabled the modified element to model concrete cracking, reinforcing bar failure and gross transverse failure. The concrete failure model used is the Hsieh-Ting-Chen four-parameter model, so that to represent the following four failure states (Kulak et al., 1997): 1. 2. 3. 4.
this case the loading from weight of the vehicle with the container, and also loading from drop of basket filled with 6 SFAs is imposed on points of contact of transportation vehicle wheels with rails (Fig. 4, loading is presented by arrows), and no loading is carried by jacks. The calculations performed for the standard mechanical characteristics of slab showed that the destructions of floor will not occur, but formation of cracks in concrete will start at loading of floor of the transport corridor. Fig. 5 shows the loading evolution of the concrete failure index (CFAIL) for each layer in most loaded element. The tensile failure surface of concrete is reached at layers 3–5, which are located in the tension side of the floor, and this is shown in Fig. 5 by CFAIL = 1 (limit for tension is reached in Layer 3, where the concrete cracking begins, and only tensile failure is reached) and CFAIL = 2 (tensile and compressive failure surface is reached—tension evaluation is completed and the crack in concrete starts to open). Layers 1–2 have CFAIL = 0 (no failure), and this indicates that neither the tensile nor compression failure surface has not been reached. The failure of the concrete begin at 400 t load at layers 3–5. The experimental (testing) mechanical characteristics of concrete obtained as a result of special investigations is much better than standard. The NEPTUNE analysis results using experimental properties of concrete and reinforcing bars material are presented in Fig. 6. The maximum normal stress in concrete is 1.37 MPa (see Fig. 6). It means that in concrete of floor ultimate strength to tension is not achieved and opening of cracks does not begin (tension, tensile strength is 3.7 MPa). The calculations performed for experimental mechanical characteristics of concrete prove that the destructions of slab will not occur, also the formation of cracks in concrete will not start at loading of floor of the transport corridor by weight of 500 tons (5 MN), i.e., floor maintains this loading.
uniaxial compressive strength, uniaxial tensile strength, equal biaxial compressive strength, combined triaxial compression.
NEPTUNE calculates stresses on central axis of element in five integration points for concrete and in each rebar layer. In order to evaluate structural integrity of slab during drop of fuel-loaded basket from a height of 17 m, conservatively an additional failure of jacks of the transportation vehicle was assumed. In
Fig. 6. Maximum normal stress (Pa) distribution in concrete of the slab.
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The strength analysis of floor of the transport corridor showed that the slab will sustain drop of fuel-loaded basket in the case with the failure of jacks. 3.3. Safety justification of container in the case of fire The performed calculations demonstrated that decay heat is efficiently removed from SFAs to the container vessel and further to the environment by the natural convection and radiation and during fuel transportation at normal operation conditions the temperature of fuel roads does not exceed 153 ◦ C. To meet the requirements of regulations, fuel roads shall not exceed the permissible temperature (300 ◦ C) not only for normal operation conditions, but also during the fire. Therefore, the safety analysis of fuel transportation in the case of fire is also provided in the safety analysis report. For thermal characteristics calculation in the case of fire, the parameters of fire are accepted in accordance with the Regulations OPBZ-83 (1983) and TS-R-1 (2009) when fire temperature is equal to 800 ◦ C, whereas duration of fire is 0.5 h. Such fire could be caused by the spilt of liquid fuel during the accident. In addition, as the fire parameters in case of container storage, flame temperature is assumed to be 600 ◦ C, whereas fire duration—1 h, like in the case of the simulation of the CONSTOR RBMK-1500 container in the case of fire. The calculation of thermal characteristics of the container is carried out using the method of finite elements employing three-dimensional programme ANSYS version 5.5.1. Designed finite-element model of the container contains 3789 finite elements and 3545 junctions. Calculation results show the available margin to the maximum permissible temperature of fuel rods in the case of fire. In case of container appearance in the fireplace with the flame temperature of 600 ◦ C, at fire duration of 1 h and with the subsequent cooling in the air, the maximum temperature of fuel rods is achieved in 4.3 h after entering the fireplace and comprises 188 ◦ C. In case the container enters the fireplace with temperature of 800 ◦ C at fire duration of 0.5 h and with the subsequent cooling in the air, the maximum temperature of fuel rods is achieved in 3.7 h after entering the fireplace and comprises 201 ◦ C. Level of temperatures of metal elements in the case of fire (approximately 200 ◦ C) and insignificant duration of such temperatures do not have a negative impact on the equipment in general. The simulation showed that maximum temperature gradients occurring in thick-walled elements of the casing in especially unfavorable time spans and due to that obtained thermal stresses do not reduce the strength of these constructive elements. Obtained calculation results confirmed that in case of transportation container entrance into fire the failure of fuel cladding integrity does not occur. 3.4. Nuclear criticality safety The original design calculations for nuclear criticality safety evaluation for spent nuclear fuel transport container were performed by employing codes MCNP and (KENO-VI). In this paper the results of verification calculations are presented and compared with design calculations. Verification calculations were performed using SCALE (USA) computer code, which uses the criticality program KENO-V (Petrie et al., 2006). In the verification calculations nuclear criticality analysis of spent nuclear fuel container was made in several stages. The basket with six fuel assemblies is located in the container. First of all the model of RBMK-1500 fuel assembly was developed (Fig. 7) and then it was included in the model of container with the basket for 6 SFAs. The RBMK-1500 fuel assembly consists of two segments (upper and lower fuel bundles), 18 fuel each. Fuel elements are arranged in two concentric rings with a central carried rod (Fig. 7).
Fig. 7. RBMK-1500 fuel assembly digital model.
The geometrical characteristics of fuel assembly are presented in the Ignalina NPP Source book (Almenas et al., 1998) and are used in the model developed by employing SCALE computer code. Spent nuclear fuel container criticality analysis was made in several stages by employing the SCALE code. First, the geometrical model of the RBMK-1500 fuel assembly was made. Then was modeled fresh fuel burning process, using SAS2H module of SCALE code. During fuel burning process a new unstable nuclides are generated. In the second stage, the obtained new fuel nuclides composition was used for the criticality calculation. The study aims to verify whether the criticality value does not exceed the allowable limits based on IAEA documents. The same methodology was used by Calic and Ravnik (2010). They carried out calculations with MCNP code. At this work were investigated 3 different configurations of spent nuclear fuel assemblies in baskets of the container (Fig. 8). First case represents normal transportation configuration, and the second with the third are hypothetical conservative cases then the spent fuel assemblies are sagged to critical mass. The criticality calculations for all three configurations were made at the different water concentrations (0.01–1.00 g/cm3 ) and fuel enrichments (2.0% and 2.8% U235 ). As shown in Fig. 9 the highest criticality values are received, then the fuel assemblies are sagged to one place (see 3rd case in Fig. 8) and the fuel enrichment is 2.8%. In this case criticality value could reach about 0.7 in the case of water density 1.00 g/cm3 . However in reality not more than 10 l of water can appear in the container during fuel loading process. This corresponds to the average water density of 0.14 g/cm3 and criticality value in this case does not exceed 0.4. The received nuclear criticality safety results were compared with the results of design calculations. Fig. 10 presents the comparison of criticality calculation results received by different computer codes and different authors for normal transportation configuration in the case of 2.8% fuel enrichment (i.e., maximal enrichment for RBMK-1500 fuel). As we can see in Fig. 10 the results of verification calculations performed by the SCALE (KENO-V) code are very similar to the results of design calculations received by KENOVI and MCNP computer codes. The difference between different results is just about 2%. Thus, the performed calculations proved that a nuclear criticality safety is assured with large margin, because the acceptance criterion keff = 0.95 is not achieved for both normal operation and accident conditions. 3.5. Radiological safety 3.5.1. Radiological safety during normal operational conditions The original design calculations of sum dose rates for spent nuclear fuel transport container were performed by employing code MCNP. In this paper the results of verification calculations
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Fig. 9. The criticality dependence from water density of the container with 6 RBMK1500 nuclear fuel assemblies (1st case, 2nd case and 3rd case are the arrangements of spent nuclear fuel assemblies in the container—see Fig. 8).
are presented and compared with design calculations. Verification calculations were performed using SCALE (USA) computer code, which uses the sum dose rate calculation module SAS 4 (Tang and Emmett Margaret, 2006). The investigation was made for the different spent nuclear fuel cooling times (1, 1.5, 3, 5 years) in the water pools, and different burnup levels (4.5, 5, 10, 15, 15.6, and 18.9 MWd/kgU). The sum dose rates were calculated on the external container surface and at different distances from this surface. The calculated sum dose includes gamma and neutrons doses. As already mentioned in Section 3.4, first of all fresh fuel burning process was modeled. New unstable nuclides are generated during the fuel burning process, and a certain amount of energy is emitted (energy depend from the fuel burnup level). The obtained amount of energy was used for dose rate calculations of spent nuclear fuel container. For dose rate calculations spent fuel container was simulated by 3 zones (Fig. 11). The central inner part of geometrical model of spent fuel container represents six RBMK-1500 fuel assemblies, which were homogenized in one zone (Zone 1 in Fig. 11). Other zone of model represents single area of air space (Zone 2 in Fig. 11) and a container wall (Zone 3). The same methodology was used by Margeanu and Angelescu (2003). The homogenous model of container with nuclear fuel was developed for dose rate assessment. Six fuel assemblies were homogenized to one zone in this model (Fig. 11).
Fig. 8. The arrangements of spent nuclear fuel assemblies in transport container baskets. Fig. 10. The transport container with 2.8% U235 fuel enrichment criticality dependence from water density (1st case the arrangements of spent nuclear fuel assemblies in the container—see Fig. 8).
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Fig. 13. Comparison of verification calculations (SCALE) and design results (MCNP). Fig. 11. Model of transport container with spent fuel: 1, homogenous fuel zone; 2, void; 3, container wall (steel).
Fig. 14. Calculation results comparison with measurements. Fig. 12. Dependence of sum dose rate on the external container surface from the spent burnup level and fuel cooling time.
The results of dose rate on the external container surface show that dose rate is largest when the cooling time is the smallest and burnup level is the highest (Fig. 12). The SCALE calculation results of dose rates on the external container surface and at different distances from this surface (0.5, 1.0, and 2.0 m) were compared with results of design calculations performed by the MCNP computer code. In accordance with boundary conditions and results of optimization calculations the spent nuclear fuel is transported by the container after at least 1.5 years cooling in the water pool, whose maximal burnup level does not exceed 18.9 MWd/kgU and initial enrichment of U235 is 2.6%. Fig. 13 presents results received for these most conservative conditions regarding nuclear fuel cooling time and burnup level. The comparison showed that calculated dose rates are very similar, the difference between SCALE and MCNP results does no exceed 5% (Fig. 13). Fig. 13 also shows that the maximal calculated dose rate value on the container external surface is 214 Sv/h for the most conservative conditions regarding nuclear fuel cooling time, enrichment and burnup level. This calculated value is almost ten times lower even than the limit of 2 mSv/h, which is defined in the International Atomic Energy Agency Safety Standard TS-R-1 (2009) as the maximum radiation level at any point on the external surface of a package for radioactive material transportation on a public roads and railways (in our case the container is used for on-site
fuel transportation). The radiation levels at a distances of 1 m and 2 m from the external surface of the package is also well below the limits prescribed by the IAEA Safety requirements No. TS-R-1 (2009). The calculation results were compared also with data of experimental measurements of dose rates on the external container surface and at the different distances from it for the fuel burnup levels of 4.5 and 15.6 MWd/kgU. Fig. 14 shows that calculation values are at ∼30% higher than measurement values. Thus, it can be concluded that calculated values are conservative and assures additional safety margin. Broadhead et al. (1997) obtained about 30% relative error of measurement comparing with calculation values. Ene (2000) pointed important errors (up to 30%) caused by modeling limitations. Thus, the other authors received similar differences between measurements and calculations as presented in this paper. 3.5.2. Radiological safety in the case of beyond design basis accident Strength analysis calculations confirmed that there is no fuel rods damage in case of drop of full-loaded basket (i.e., with 6 SFAs) from the maximum possible height (17 m), even in the case of single failure of one of the elements of basket brake device (i.e., in the case of design basis accident with single failure). For the assessment of radiological consequences the beyond design basis accident scenario – drop of full-loaded basket from the maximum possible height with simultaneous complete failure of basket brake device (simultaneous failure of all brake blocks) – was analyzed. It was
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assumed that in this scenario the loss of leak tightness of all fuel rods in all 6 SFAs occurs. Thus release of radioactive nuclides located in gas spaces under fuel rod claddings will occur. The carried out assessment of radiological consequences on the border of INPP controlled area (3 km) has shown that as a result of analyzed accident scenario the radiation dose for the population (taking into account all ways of an irradiation) for the first year after the accident will not exceed 11 Sv, that is significantly below than the annual effective limited dose of the population for operation conditions—200 Sv (HN 87, 2002) and, thus, intervention on the protection of the population is not required in accordance with regulations (HN 73, 2001). 4. Conclusions The safety analysis report demonstrated that the proposed transportation container and other equipment, which are intended for on-site transportation of RBMK reactor fuel, are in compliance with functional, design and regulation requirements and assure the required safety level for both normal operation and accident conditions. The calculation demonstrated that maximum permissible temperature of fuel rods (300 ◦ C) is not achieved even in the case of fire (maximal temperature is ∼200 ◦ C). Results of the calculations show that all constructive elements (spent fuel assembly, the container) maintain their structural integrity even in the case of dropping a fuel-loaded basket from the maximal possible height (17 m). Radiological and nuclear safety criteria are met for both normal operation and accident conditions. The developed equipment can be used also in decommissioning phase for fuel transportation to fuel storage facilities. The set of this equipment can be applied for other NPPs with RBMK type reactors. Acknowledgments The authors want to extend thanks to the administration and technical staff at the Ignalina NPP, for providing information regarding operational procedures and operational data.
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