Mechanical safety analysis for high burn-up spent fuel assemblies under accident transport conditions

Abstract

Transport packages for spent fuel have to meet the requirements concerning containment, shielding and criticality as specified in the International Atomic Energy Agency regulations for different transport conditions. Physical state of spent fuel and fuel rod cladding as well as geometric configuration of fuel assemblies are, among others, important inputs for the evaluation of correspondent package capabilities under these conditions. The kind, accuracy and completeness of such information depend upon purpose of the specific problem. In this paper, the mechanical behaviour of spent fuel assemblies under accident conditions of transport will be analysed with regard to assumptions to be used in the criticality safety analysis. In particular the potential rearrangement of the fissile content within the package cavity, including the amount of the fuel released from broken rods has to be properly considered in these assumptions. In view of the complexity of interactions between the fuel rods of each fuel assembly among themselves as well as between fuel assemblies, basket, and cask body or cask lid, the exact mechanical analysis of such phenomena under drop test conditions is nearly impossible. The application of sophisticated numerical models requires extensive experimental data for model verification, which are in general not available. The gaps in information concerning the material properties of cladding and pellets, especially for the high burn-up fuel, make the analysis more complicated additionally. In this context a simplified analytical methodology for conservative estimation of fuel rod failures and spent fuel release is described. This methodology is based on experiences of BAM acting as the responsible German authority within safety assessment of packages for transport of spent fuel.     

Multiple water barriers: alternative to assessment of fuel assemblies during accident conditions of transport

Abstract

Packages for the transport of radioactive material have to comply with national and/or international regulations. These regulations are widely based on the requirements set forth by the International Atomic Energy Agency (IAEA) in the 'Regulations for the safe transport of radioactive material'. In this framework, packages to transport fuel assemblies (including spent fuel assemblies) have to meet the requirements for packages containing fissile material. In accident conditions of transport, the applicant for the package design approval has to show that the package remains subcritical taking due account of the status of the contents in these conditions. In most cases, considering water ingress in the package, it is not possible to assume that the fissile material included in the fuel assemblies is dispersed in the package with the most severe conceivable distribution regarding criticality. In order to alleviate this difficulty, during the last years, we have provided a significant better knowledge of the conditions of the fuel assemblies to be transported. This was part of the Fuel Integrity Project, whose progress was regularly reported during PATRAM 2001 and PATRAM 2004 Symposia. However, for packages which encounter a large g-load during accident conditions of transport and/or which contain spent fuel assemblies with very high burn-up, it can be difficult to demonstrate that the fuel assemblies are not significantly damaged. Then, to make the criticality assessment considering water inleakage into the flask and a large release of fissile material within its cavity will not allow meeting the subcriticality criteria. For that reason, for our package designs, which use a gas and not water as an internal coolant and which fall into that category, the author has decided to take credit of the possibilities provided by the subparagraph 677 (b) of the Regulations. This paragraph allows not taking into account water in the package, provided that the package exhibits 'multiple high standard water barriers'. The paper describes the author's experience with the implementation of this paragraph. Two different cases are considered: either a double vessel, or a double lid. It will be explained when each of these solutions is implemented, and give examples of package designs with such features, as well as the approvals which were granted for these designs in various countries.     

New approach to evaluate lattice expansion of light water reactor fuel elements on criticality safety of transport packages under impact accidents

Abstract

In order to safely transport packages containing light water reactor fuel assemblies, it is essential to maintain the fuel assemblies in a subcritical state in accidents during transport. To evaluate nuclear criticality safety, an estimator is required to determine an absolutely safe level based not only on hypothetical accidents but also on practical accident levels which, to some extent, are based on actual accidents. The purpose of the present study is to suggest the arrangement of the deformation range of the fuel assembly after an actual accident, and to obtain the maximum value of the neutron effective multiplication factor based on the criticality safety assessment for the transport cask. In the present study, two kinds of criticality calculations for the package were considered: large scale pin pitch shift and small scale pin pitch shift. For the large scale pin pitch shift, a parameter which determines the location of each fuel pin which constitutes the fuel assembly was introduced so that the criticality calculation for the fuel assembly with non-uniform lattice pitch can be performed parametrically. The result of the criticality calculation using the parameter made it clear that the fuel pin pitch is sensitive to the neutron reactivity because each of the fuel pin pitches is related to a ratio of the fissile to the moderator, and that the relationship of the ratio to the neutron reactivity depends on the type of the fuel assembly involved, i.e. the type of a nuclear reactor in which a fuel assembly is used. For the small scale pin pitch shift, the study focused on the small displacement of each fuel pin. The small displacement of each fuel pin pitch can be described probabilistically using the stochastic geometry routine in MCNP code. Using the scheme in combination with the scheme for the large scale pin pitch shift, the maximum value of the neutron effective multiplication factor of the package after an accident can be obtained. This scheme is useful to determine the maximum neutron effective multiplication factor for the criticality safety evaluation.     

Transport of MOX fuel: continuous challenge

Abstract

For 45 years TN International has been involved in the radioactive materials transportation field. Since the beginning the spent nuclear fuel transportation has been its core business. During all these years TN International, now part of AREVA, has been able to anticipate and fulfil the needs for new transport or storage casks design to fit the nuclear industry evolutions. A whole fleet of casks able to transport all the materials of the nuclear fuel cycle has been developed. This paper focuses on the casks used to transport the fresh and used mix oxide (MOX) fuel. To transport the fresh MOX boiling water reactor and pressurised water reactors fuel, TN International has developed two designs of casks: the MX 6 and the MX 8. These casks are and have been used to transport MOX fuel for French, German, Swiss and in a near future Japanese nuclear power plants. A complete set of baskets have been developed to optimise the loading in terms of integrated dose and also of course capacity. Mixed oxide used fuel has now its dedicated cask: the TN 112 which certificate of approval has been obtained in July 2008. This cask is able to transport 12 MOX spent fuel elements with a short cooling time. The first loading of the cask has been performed in September 2008 in the Electricité de France nuclear power plant of Saint-Laurent-des-Eaux. By its continuous involvement in the nuclear transportation field, TN International has been able to face the many challenges linked to the radioactive materials transportation especially talking of MOX fuel. TN International will also have to face the increasing demand linked to the nuclear renaissance.     

BWR spent fuel transport and storage with TN9/4 and TN24BH casks

Abstract

In 2001 the Swiss nuclear utilities started to store spent fuel in dry metallic dual purpose casks at ZWILAG, the Swiss interim storage facility. BKW FMB Energy Ltd, as the owner of the Mühleberg nuclear power plant, is involved in this process and has selected to store the spent fuel in a new high capacity dual purpose cask, the TN24BH. For the transport Cogema Logistics has developed a new medium size cask, the TN9/4, to replace the NTL9 cask, which has performed numerous shipments of BWR spent fuel in past decades. Licensed by the IAEA 1996, the TN9/4 is a 40 t transport cask, for seven BWR high burnup spent fuel assemblies. The spent fuel assemblies can be transferred to the ZWILAG hot cell in the TN24BH cask. These casks were first used in 2003. Ten TN9/4 shipments were made, and one TN24BH was loaded. After a brief presentation of the operational aspects, the paper will focus on the TN24BH high capacity dual purpose cask and the TN9/4 transport cask and describe in detail their characteristics and possibilities.     

A methodology for the evaluation of fuel rod failures under accident conditions

Abstract

Recent studies on the long-term behaviour of high-burnup spent fuel have shown that, under normal conditions of storage, challenges to cladding integrity from various postulated damage mechanisms, such as delayed hydride cracking, stress-corrosion cracking and long-term creep, would not lead to any significant safety concerns during dry storage, and regulatory rules have subsequently been established to ensure that a compatible level of safety is maintained. However, similar regulatory rules have not yet been developed to address failures of fuel rod cladding that could potentially lead to reconfigured fuel geometry under hypothetical transport accidents. At issue is the effect on cladding ductility of potential changes in zirconium hydride morphology during dry storage. Recent studies have shown that above a certain level of cladding hoop stress, the decaying temperature history during dry storage can cause the hydrogen in solid solution to precipitate in the form of radial hydrides, which, depending on their relative concentration, can induce brittle failures in the cladding. From a US regulatory perspective such cladding failures, if they were to cause fuel reconfiguration, could invalidate the cask's criticality and shielding licensing analyses, which are based on coherent geometry. This paper describes a methodology for high-burnup spent fuel to determine the frequency of cladding failure and failure modes under drop accidents, considering end-of-storage spent fuel conditions. The degree to which spent fuel reconfiguration could occur during handling or transport accidents would depend to a large extent on the number of fuel rod failures and the type and geometry of the failure modes. Such information can only be developed analytically, as there are no direct experimental data that can provide guidance on the level of damage that can be expected. To this end, this paper focuses on the development of a methodology for modelling and analysis that deals with this general problem on a generic basis. First, consideration is given to defining accident loading that is equivalent to the bounding hypothetical transport accident of a 9 m drop onto an essentially unyielding surface. Second, an analytically robust material constitutive model, an essential element in a successful structural analysis, is required. A model of material behaviour, with embedded failure criteria, for cladding containing various concentrations of circumferentially and radially oriented hydrides has been developed and implemented in a finite-element code. The hydride precipitation model, which describes the hydride structure of the cladding at the end of dry storage, and the hydride-dependent properties of high-burnup fuel cladding form the main input to the constitutive model. The third element in the overall process is to utilise this material model and its host finite-element code in the structural analysis of a transport cask subjected to bounding accident loading to calculate fuel rod failures and failure mode configurations. This requires detailed modelling of the transport cask and its internal structure, which includes the canister, basket, fuel assembly grids and fuel rods. The overall methodology is described.     

Spent fuel transportation risk assessment: conclusions

Abstract

The US Nuclear Regulatory Commission (NRC) has recently completed an updated Spent Fuel Transportation Risk Assessment, NUREG-2125. The study reached the following findings. First, the collective dose risks from routine transportation are vanishingly small. These doses are about four to five orders of magnitude less than collective background radiation doses. Second, the routes selected for this study adequately represent the routes for spent nuclear fuel transport, and there was relatively little variation in the risks per kilometre over these routes. Third, radioactive material would not be released in an accident if the fuel is contained in an inner welded canister inside the cask. Fourth, only rail casks without inner welded canisters would release radioactive material, and only then in exceptionally severe accidents. Fifth, if there were an accident during a spent fuel shipment, there is less than one in a billion chance the accident would result in a release of radioactive material. Sixth, if there were a release of radioactive material in a spent fuel shipment accident, the dose to the maximally exposed individual would be <2 Sv (200 rem) and would not cause an acute fatality. Seventh, the collective dose risks for the two types of extraregulatory accidents (accidents involving a release of radioactive material and loss of lead shielding) are negligible compared to the risk from a no release, no loss of shielding accident. Eight, the risk of loss of shielding from a fire is negligible. Ninth, none of the fire accidents investigated in this study resulted in a release of radioactive material. Based on these findings, this study reconfirms that radiological impacts from spent fuel transportation conducted in compliance with NRC regulations are low. In fact, this study’s radiological impact estimates are generally less than the already low estimates reported in earlier studies. Accordingly, with respect to spent fuel transportation, this study reconfirms the previous NRC conclusion that the regulations for transportation of radioactive material are adequate to protect the public against unreasonable risk.     

Mechanical characteristics of fuel rod claddings in transport conditions

Abstract

A synthesis on the mechanical characteristics of unirradiated and irradiated fuel rod claddings was performed by the French Institut de Radioprotection et de Sûreté Nucléaire (IRSN) in order to have reference data for the assessment of the safety demonstrations in normal and accident conditions of transport required by the procedure of package licensing. Indeed, the transport conditions correspond to a range of cladding temperatures (200–550°C) which is only partly covered by the data acquired within the framework of the safety demonstration relative to the reactor normal operating conditions, especially beyond 400°C. This work concerned Zircaloy-4 cladding material (Zry-4) and M5^TM. Data about mechanical properties (elastic and ductile properties, creep behaviour), oxidation (in reactor and under air during transport), hydrides and fracture toughness have been collected and synthesised. The laws presented in the document can be used to obtain orders of magnitude of oxide layer thickness, hydrogen content and creep deformation rate. The following phenomena which could influence the mechanical behaviour of the cladding were more particularly studied: oxidation which could become very important during transport in case of cladding temperatures of ～500°C; creep for which only a few data ～500°C are available and which depends in particular on the internal pressure of the rods, the cladding oxidation and the presence of the hydrides; and recrystallisation of Zry-4 at ～500°C, which could have consequences on the mechanical properties of the cladding after cooling during the storage. For other topics of interest for the study of the mechanical behaviour of the cladding, such as the fracture toughness for example, it was identified that the data available is scarce.     

Spent fuel transport using cask loaded outside pool

Abstract

During the last year, Sogin (the Italian company in charge for decommissioning of Italian nuclear power plants) had to implement an accelerated decommissioning plan of a EUREX spent fuel pool due to finding a water leakage into the environment from the pool. EUREX is no longer operating a pilot reprocessing plant, which some years ago became the responsibility of Sogin. There were 52 spent fuel assemblies from the Trino Vercellese PWR nuclear power plant, 48 irradiated pins from a Garigliano BWR fuel assembly, and 10 plates from an irradiated MTR fuel assembly stored in the EUREX pool, so the first step of the accelerated decommissioning plan consisted in the evacuation of this spent fuel. Considering the necessity to start the evacuation as soon as possible, Sogin decided to use an already existing cask (AGN-1) used in the past for the transport of Trino and Garigliano fuel assemblies. This cask was requalified in order to obtain a transport licence for the fuel assemblies stored in the EUREX pool according to ADR 2005 regulation. The transport license for the AGN-1 cask loaded with EUREX fuel assemblies was released by APAT (the Italian Safety Authority) in the spring of 2007. Owing to the limited capacity of the EUREX pool crane (27 t for nuclear loads) and limited dimensions of pool operational area, it was not possible to transfer the AGN-1 cask (50 t) into the pool for fuel assemblies charging. The solution implemented to overcome this problem was the loading of the cask outside the pool. A special shielding shuttle was developed and used to allow safe spent fuel transfer between the pool and the cask. This procedure avoided also the problem of excessive contamination of cask surfaces that could have occurred due to very high level of contamination of EUREX pool water if the cask had been immersed in the pool. Additional shielding devices were developed and used to reduce dose rate during cask loading operations. Although the evacuation of spent fuel assemblies from the EUREX pool was a very challenging activity due to the short time available, unfavourable space conditions inside the pool building and handling tool limitations; all loading and transport operations were performed successfully and without particular problems. Ten transports were carried out to evacuate all of the spent fuel stored in the EUREX pool. Spent fuel was transferred to the Avogadro Deposit pool. The first loading sequence started on 2 May 2007 and the first transport was performed on 6 May 2007. The tenth and last transport was performed on 21 July 2007. A dose less than 50 μSv (neutron + gamma) was measured for the most exposed operator during a complete cask loading sequence.     

Analyses of historical rail accidents to identify accident parameters impacting transport of spent nuclear fuel

Abstract

As a regulatory authority for the transportation of spent nuclear fuel (SNF) in the USA, the Nuclear Regulatory Commission requires that SNF transportation packages be designed to endure a fully engulfing fire with an average temperature of 800°C (1475°F) for 30 min, as prescribed in Title 10 of the Code of Federal Regulations Part 71. The work described in this paper was performed to support the Nuclear Regulatory Commission in determining the types of accident parameters that could produce a severe fire with the potential to fully engulf an SNF transportation package. This paper describes the process that was used to characterise the important features of rail accidents that would potentially lead to an SNF transport package being involved in a severe fire. Historical rail accidents involving all hazardous material (i.e. all nine classes of hazardous material) and long duration fires in the USA have been analysed using data from the Federal Railroad Administration and the Pipeline and Hazardous Materials Safety Administration. Parameters that were evaluated from these data include, but were not limited to, class of track where the accident occurred, class of hazardous material that was being transported and number of railcars involved in the fire. The data analysis revealed that in the past 34 years of rail transport, roughly 1800 accidents have led to the release of hazardous materials, resulting in a frequency of roughly one accident per 10 million freight train miles (Because all of the data were obtained in the USA, which still uses distance measured in miles, and the primary source is an extensive database from the Federal Railroad Administration that is also in reported in miles, the data in this paper are reported in miles rather than kilometres. Conversion of miles to kilometres is by multiplication of 1·61.). In the last 12 years, there have only been 20 accidents involving multiple car hazardous material releases that led to a fire. This results in an accident rate of 0·003 accidents per million freight train miles that involved multiple car releases and a fire. Out of all the accidents analysed, only one involved a railcar carrying class 7 (i.e. radioactive) hazardous material.     

Vulnerabilities of legacy fuel casks when evaluated using modern structural analysis tools

Abstract

Cylindrical fuel casks often have impact limiters surrounding the ends of the cask shaft in a typical 'dumbbell' arrangement. The primary purpose of these impact limiters is to absorb energy to reduce loads on the cask structure during impacts associated with a severe accident. Impact limiters are also credited in many packages with protecting closure seals and reducing peak temperatures during fire events. For this credit to be taken in safety analyses, the impact limiter attachment system must be shown to retain the impact limiter following normal conditions of transport (NCT) and hypothetical accident conditions (HAC) impacts. Large casks are often certified by analysis only because of the cost associated with testing. Therefore, some cask impact limiter attachment systems have not been tested in real impacts. A recent structural analysis of the T-3 spent fuel containment cask found problems with the design of the impact limiter attachment system. Assumptions in the original safety analysis for packaging (SARP) concerning the loading in the attachment bolts were found to be inaccurate in certain drop orientations. This paper documents the lessons learned and their applicability to impact limiter attachment system designs.     

Fundamental safety strategy against severe accidents on prototype sodium-cooled fast reactor

The accident categories of severe accidents (SAs) for prototype sodium-cooled fast reactor (SFR) which need proper measures were investigated through the internal event probabilistic risk assessment (PRA) and event tree analysis for the external event and six accident categories, unprotected loss of flow (ULOF), unprotected transient over power (UTOP), unprotected loss of heat sink (ULOHS), loss of reactor sodium level (LORL), protected loss of heat sink (PLOHS) and station blackout (SBO), were identified. Fundamental safety strategy against these accidents is studied and clearly stated considering the characteristics and existing accident measures of prototype SFR, and concrete measures based on this safety strategy are investigated and organized. The sufficiency of these SA measures is confirmed by comparing the evaluated core damage frequency (CDF) and containment failure frequency (CFF) to the target value, 1×10^?5 and 1×10^?6 per plant operating year, respectively, which were selected based on the IAEA's safety target. However, the target value of CDF and CFF should be satisfied considering all the SAs caused by both internal and external events. External event PRA for prototype SFR is now under evaluation and we set out to satisfy the target value of CDF and CFF considering both internal and external events.     

Transport of radioactive materials: the need for radiation protection programmes

Abstract

The increase in the use of radioactive materials worldwide requires that these materials be moved from production sites to the end user, or in the case of radioactive waste, from the waste generator to the repository. Tens of millions of packages containing radioactive material are consigned for transport each year throughout the world. The amount of radioactive material in these packages varies from negligible quantities in shipments of consumer products to very large quantities in shipments of irradiated nuclear fuel. Transport is the main way in which the radioactive materials being moved get into the public domain. The public is generally unaware of the lurking danger when transporting these hazardous goods. Thus radiation protection programmes are important to assure the public of the certainty of their safety during conveyance of these materials. Radioactive material is transported by land (road and rail), inland waterways, sea/ocean and air. These modes of transport are regulated by international 'modal' regulations. The international community has formulated controls to reduce the number of accidents and mitigate their consequences should they happen. When accidents involving the transport of radioactive material occur, it could result in injury, loss of life and pollution of the environment. In order to ensure the safety of people, property and the environment, national and international transport regulations have been developed. The appropriate authorities in each state utilise them to control the transport of radioactive material. Stringent measures are required in these regulations to ensure adequate containment, shielding and the prevention of criticality in all spheres of transport, i.e.routine, minor incidents and accident conditions. Despite the extensive application of these stringent safety controls, transport accidents involving packages containing radioactive material have occurred and will continue to occur. When a transport accident occurs, it is unlikely to result in a significant release of radioactive material, loss of shielding or loss of criticality control.     

Energy and Angular Responses of the Criticality Accident Detector using a Plastic Scintillator

The Japan Atomic Energy Agency (JAEA), Nuclear Fuel Cycle Engineering Laboratories, operates a spent fuel reprocessing plant and MOX (Plutonium-Uranium Mixed Oxide) fuel fabrication plants. Criticality accident detectors have been installed in these facilities. The detector, the Toshiba RD120, is composed of a plastic scintillator coupled to a photomultiplier tube, and an operational amplifier. The alarm triggering point is set to 1.0–3.6 mGy.h^?1 in photon dose rate to detect the minimum accident of concern. However, a plastic scintillator is principally sensitive not only to primary photons but also to neutrons by secondary photons and heavy charged particles produced in the detector itself. The authors calculated energy and angular responses of the RD120 criticality accident detector to photons and neutrons using Monte Carlo computer codes. The response to primary photons was evaluated with the MCNP-4B and EGS4 calculations, and photon and X-ray irradiation experiments. The response to neutrons that produce secondary photons and heavy charged particles from neutron interactions was computed using the MCNP-4B and SCINFUL, respectively. As a result, reliable response functions were obtained. These results will be a great help in reassessing the coverage area and in determining the appropriate triggering dose rate level in criticality accidents.     

Emergency Preparedness and Response in Case of a Fire Accident with UF6 Packages Traversing the Suez Canal

Abstract

Egypt has a unique problem: the Suez Canal. Radioactive cargo passes regularly through the canal carrying new and spent reactor fuel. There are also about 1000 metric tonnes of uranium hexafluoride (UF₆) passing through the canal every year. In spite of all the precautions taken in the transport, accidents with packages containing UF₆ shipped through the Suez Canal may arise, even though the probability is minimal. Such accidents may be accompanied by injuries to or death of persons and damage to property including radiation and criticality hazards and high chemical toxicity, particularly if the accident occurred close to one of the three densely populated cities (Port Said, Ismailia and Suez), which are located along the west bank of the Suez Canal. The government of Egypt has established a national radiological emergency plan in order to deal with any radiological accidents which may arise inside the country. This paper considers the effect of a fire accident to industrial packages containing UF₆ on board a cargo ship passing along the Suez Canal near Port Said City. The accident scenario and emergency response actions taken during the different phases of the accident are presented and discussed. The paper highlights the importance of public awareness for populations located in densely populated areas along the bank of the Suez Canal, in order to react in a timely and effective way to avoid the toxic and radiological hazards resulting from such a type of accident. The possibility of upgrading the capabilities of civil defence and fire-fighting personnel is also discussed.     

Comparison Of Results From Two Spent Fuel Sabotage Source Term Experiments

Abstract

Estimation of the radiological impact of a sabotage attack on a spent fuel cask requires values for the respirable spent fuel aerosol that might be produced by the attack. Two potential sources of relevant source term data are reviewed and shown to provide consistent information which leads to values of 6 × 10^?5 to 8 × 10^?4 g of respirable surrogate spent fuel aerosol released from the cask per gram of surrogate fuel matrix disrupted by a sabotage attack using a high energy density device (HEDD). Using a respirable spent fuel to respirable surrogate conversion factor of 3 enables estimation of the spent fuel respirable source term from the experiments reviewed.     

Potential effects of recent transportation accident on transportation of radioactive materials in USA

Abstract

In 2007, two severe transportation accidents, involving primarily long haul tractor trailers, occurred in the State of California. In the first, which occurred in Oakland in what is commonly known as the 'MacArthur Maze' section of Interstate 580 (I-580), a tractor trailer carrying gasoline impacted an overpass support column and burst into flames. The subsequent fire burned for over 2 h and led to the collapse of the overpass due to the loss of strength in the structural steel that supported the overpass. The second incident was a chain reaction accident involving several tractor trailers in the I-5 'Newhall Pass' truck bypass tunnel in Santa Clarita. This accident also involved an intense fire that damaged the tunnel and required the closing of the tunnel for repairs to the concrete walls. The US Nuclear Regulatory Commission (NRC) is studying both these accidents to examine any potential regulatory implications related to the safe transport of radioactive materials including spent nuclear fuel in the United States. This paper will discuss the details of the MacArthur Maze fire investigation. The NRC's investigation of the Newhall Pass fire is ongoing and the results are not available for publication.     

Accident Risk Assessment of Transport of Spent Fuel by Rail in India

Abstract

The results are presented of a study carried out to estimate the accident risk involved in the transport of spent fuel from a typical Indian nuclear power plant site near Kota to a typical fuel reprocessing plant site at Tarapur. The fuel considered is the low burnup Indian pressurised heavy water reactor fuel with a minimum cooling period of 485 days. The spent fuel is transported in a cuboidal, naturally cooled shipping cask over a distance of 822 km by rail. The accident risk analysis is carried out using the fault tree methodology. RADTRAN 4 computer code is used for estimating the consequences and the accident risk. Results of the analysis indicate that the accident risk values are acceptably low and do not constitute significant additional risk to the population along the route.