Applications of the multidimensional equations to complex fuel assembly problems

For the analysis of reactors with complex fuel assemblies or fine mesh applications as pin by pin neutron flux reconstruction, the usual approximation of the neutron transport equation by the multigroup diffusion equation does not provide good results. A classical approach to solve the neutron transport equation is to apply the spherical harmonics method obtaining a finite approximation known as the P_L equations. In this line, a nodal collocation method for the discretization of these equations on a rectangular mesh is used in this paper to analyse reactors with MOX fuel assemblies. Although the 3D P_L nodal collocation method becomes feasible due to the improvements in computer hardware, a complete treatment of the detailed structure of the fuel assemblies in actual three-dimensional geometry is still prohibitive, thus, an assembly homogenization method is necessary. A homogenization method compatible with our multidimensional P_L code is proposed and tested performing heterogeneous and homogenized calculations. In this work, we apply the method to 2D complex fuel assembly configurations.     

Letter to the Editor

Abstract

Transport of fresh MOX fuel assemblies for the prototype FBR MONJU initial core started in July 1992 and ended in March 1994. As many as 205 fresh MOX fuel assemblies (109 assemblies for an inner core, 91 assemblies for an outer core and 5 assemblies for testing) were transported in nine transport missions. The packaging for fuel assemblies, which has shielding and shock absorbing material inside, meets IAEA regulatory requirements for Type B(U) packaging including hypothetical accident conditions such as the 9 m drop test, fire test, etc. Moreover, this packaging design features such advanced technologies as high performance neutron shielding material and an automatic hold-down mechanism for the fuel assemblies. Every effort was made to carry out safe transport in conjunction with the cooperation of every competent organisation. This effort includes establishment of the transport control centre, communication training, and accompanying of the radiation monitoring expert. No transport accident occurred during the transport and all the transport missions were successfully completed on schedule.     

Transport of Fresh MOX Fuel Assemblies for the MONJU Initial Core

Abstract

Transport of fresh MOX fuel assemblies for the prototype FBR MONJU initial core started in July 1992 and ended in March 1994. As many as 205 fresh MOX fuel assemblies (109 assemblies for an inner core, 91 assemblies for an outer core and 5 assemblies for testing) were transported in nine transport missions. The packaging for fuel assemblies, which has shielding and shock absorbing material inside, meets IAEA regulatory requirements for Type B(U) packaging including hypothetical accident conditions such as the 9 m drop test, fire test, etc. Moreover, this packaging design features such advanced technologies as high performance neutron shielding material and an automatic hold-down mechanism for the fuel assemblies. Every effort was made to carry out safe transport in conjunction with the cooperation of every competent organisation. This effort includes establishment of, the transport control centre, communication training, and accompanying the radiation monitoring expert. No transport accident occurred during the transport and all the transport missions were successfully completed on schedule.     

Using ‘Component Multiplication’ in Monk to Reduce Pessimism in the Dose Rate Assessment for Water-Filled (Ullaged) Transport Packages

Abstract

The external dose rates from spent fuel packages consist of gamma ray and neutron components. The source of gamma rays is from fission products and actinides in the spent fuel and from activation products in structural components of the fuel element. Neutrons originate from spontaneous fission in actinides (for example from curium isotopes) within the spent fuel and from (alpha, n) reactions in oxide fuel. However, a significant number of neutrons are produced due to further fission within the fuel. This is known as neutron enhancement or multiplication (M). To treat the effects of enhancement, the neutron source may be scaled within the dose rate calculation. In a wet package, it has been customary to determine k_effective (k_eff) for a completely water-filled package or a package with a defined water level (for the horizontal transport condition). The irradiation of the fuel is normally taken into account in calculating k_eff for this purpose. The neutron enhancement is then obtained by calculating M = 1/(1 ? k_eff)) which is then applied as a source scaling factor throughout each fuel assembly. In a wet package, there is normally an ullage volume above the water level, the package only being partially flooded. The ullage volume is designed to accommodate pressure build-up within the package. Typically the top row of fuel assemblies may be partially covered and partially uncovered by water. When the above value of M is used for fuel within the dry part of the package, dose rates above the package tend to be overestimated and can limit the carrying capability of the package. (Also, a single value of M will tend to over-predict dose rate conuibutions from all assemblies around the periphery.) Use of component multiplication (a new feature available in the MONK computer code) enables two separate values of ‘k^eff’ to be determined for the wet and dry parts of the package. These typically differ by a factor of three, leading to differences in the enhancement, M. Use of different enhancement of neutrons in the wet and dry fuel regions enables a more realistic estimate of the dose rate above the ullage to be made and could remove considerable pessimism in future applications for package approval. In a package in which the neutron component dominates, this can represent a considerable reduction in dose rate. Consequently, a fuel with greater burn-up or shorter cooling may be carried without affecting the safety of the package. To date, the technique has only been shown to be applicable to irradiated fuel. Examples of the degree of reduction in neutron dose rate are given for two typical packages.     

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.     

Thermal analysis of proposed transport cask for three advanced burner reactor used fuel assemblies

Abstract

Preliminary studies of used fuel generated in the US Department of Energy's Advanced Fuel Cycle Initiative have indicated that current used fuel transport casks may be insufficient for the transportation of said fuel. This work considers transport of three 5-year-cooled oxide advanced burner reactor used fuel assemblies with a burn-up of 160 MWD kg^–1. A transport cask designed to carry these assemblies is proposed. This design employs a 7-cm-thick lead gamma shield and a 20-cm-thick NS-4-FR composite neutron shield. The temperature profile within the cask, from its centre to its exterior surface, is determined by two-dimensional computational fluid dynamics simulations of conduction, convection and radiation within the cask. Simulations are performed for a cask with a smooth external surface and various neutron shield thicknesses. Separate simulations are performed for a cask with a corrugated external surface and a neutron shield thickness that satisfies shielding constraints. Resulting temperature profiles indicate that a three-assembly cask with a smooth external surface will meet fuel cladding temperature requirements but will cause outer surface temperatures to exceed the regulatory limit. A cask with a corrugated external surface will not exceed the limits for both the fuel cladding and outer surface temperatures.     

Criticality assessments using polyurethane foam

Abstract

Rolls-Royce has designed a package to transport and store fresh fuel assemblies and anticipates approval from the regulators for the new package design in the near future. The space between the inner and outer steel shells is filled with shaped blocks of rigid polyurethane foam, of two different densities. The criticality safety case for the fresh fuel package had to consider single packages and arrays of packages under routine, normal and accident conditions. IAEA regulatory requirements state that the criticality assessment must include investigations on the effect on the neutron multiplication factor k _eff due to impacts, flooding and fire. Sensitivity studies must also be carried out to determine the effects on the k _eff due to any uncertainties in the composition of the fuel and container materials. An important part of the criticality safety case is the treatment of the foam. The approach adopted to model the polyurethane foam is the subject of this paper. The following were investigated: (1) the effect on the k _eff of varying the elemental composition of the foam, including the removal of hydrogen; (2) the experimental analysis of burnt foam; (3) the effect of addition of water to the foam to simulate water absorption; (4) a simple representation of crushed foam to simulate knock-back in the package; (5) extreme representations of burnt foam, such as replacing foam with solid carbon or as randomly distributed spheres of carbon to represent soot. These investigations were most informative and should be considered in any criticality assessments of transport packages containing large amounts of foam in the future.     

Development and verification of a nodal approach for solving the multigroup SP3 equations

The core model DYN3D which has been developed for three-dimensional analyses of steady states and transients in thermal reactors with quadratic or hexagonal fuel assemblies is based on nodal methods for the solution of the two-group neutron diffusion equation. Loading cores with higher content of MOX fuel, the increase of the fuel cycle length, and the consideration of new reactor types are challenging for these standard methods. A nodal expansion method for solving the equations of the simplified P₃ (SP₃) approximation of the multigroup transport equation was developed to improve the accuracy of the DYN3D code. The method described in the paper is verified with pinwise calculations of a steady state of the OECD/NEA and US NRC PWR MOX/UO₂ Core Transient Benchmark. The used 16-group cross section library was generated for DORT calculations with homogenized pin cells. Two different approximations of the diffusion coefficient which occurs in the within-group form of the SP₃ equations are investigated. Using the transport cross section for the calculation of the diffusion coefficient gives much better results than those obtained with the removal cross section. The improvement of the results in comparison to a pinwise diffusion calculation is shown. The results are compared with the DORT and the heterogeneous reference solution of the code DeCART. Concerning the SP₃ calculation using the diffusion coefficient based on the transport cross section (DYN3D-SP3-TR) the deviations of the eigenvalue k_eff and the assembly powers from the transport solutions of DORT and DeCART are in the same order as those between the two transport solutions themselves. The improvement of the DYN3D-SP3-TR results in comparison to the diffusion calculation is presented. As the DYN3D-SP3-TR and DORT calculations are performed with homogenized pin cells, the pin powers of the two calculations are closer to each other than to the pin powers of the DeCART solution. To estimate the contribution of higher flux-moments, the ratio of the second flux moment to the zeroth flux moment is investigated along a horizontal row of pins. It is shown that it is low in the fuel assemblies with small peaks at the water channels. Considering the baffle and the water reflector regions higher values are obtained in the fast neutron groups.     

The Transport of Front-end Materials: Trends and Challenges for the Next Decade

Abstract

For several decades all the front end nuclear industry involved in transport operations around the world relied mainly on packagings developed in the 1970s in the United States of America and used everywhere by all means of transport to the satisfaction of the authorities. Recently, the industry has been confronted with licensing problems of some 30B overpacks as well as for packagings used for the transport of fresh PWR and BWR fuel assemblies and UO₂ powder, etc. The upgrading of existing regulations may affect the present conditions of utilisation of the 48Y cylinders for the transport of natural UF₆. This paper details the actions which have been taken by the industry to improve this situation in order to continue to perform the transport of fresh fuel material smoothly.     

Description of the German Concept of Final Storage Cask

Abstract

The German storage concept for the direct final storage of spent fuel assemblies from LWR reactors is described. The final storage concept is designed in such a way that it encompasses the whole spectrum of fuel elements to be stored from German reactors, Le. U fuel assemblies and MOX fuel assemblies with a mean bumup of 55 GW.d.t^?1 heavy metal were considered. The further design requirements are defined in such a way that the cask concept satisfies the conditions for type B(U) transport, interim storage and fmal storage. The safe long-term containment of the activity is guaranteed by an inner cask welded leak-tight; the sufficient shielding and the transport packaging are ensured by a shielding cask.     

Active reduction of the end effect by local installation of neutron absorbers

In the burnup credit analyses of interim or long-term spent fuel (SF) storage facilities and transport casks, when the average burnup value is greater than approximately 30 GWd/t, the neutron multiplication factor becomes greater if we consider the axial burnup distribution of the spent fuel assembly rather than assuming an average burnup. This phenomenon is called the “end effect” and it is one of the main technical issues in burnup credit research. The end effect is characterized by an increase of the neutron flux around the end regions of the spent fuel assemblies in the criticality calculation. However, such increase of the neutron flux has not been observed in experiments using actual spent fuel assemblies.     

Towards the development of upper subcriticality limits on the basis of benchmark criticality calculations

This paper concerns the assessment of standard point-wise neutron data libraries for criticality safety evaluations in units of the effective neutron multiplication factor, k_eff, the aim being to establish a methodology for the analysis of storage pools containing fuel assemblies discharged from the Swiss Light Water Reactors. The selected approach is based on using the Monte Carlo code MCNPX (version 2.4.0 was applied in the study at hand) and a modern standard point-wise neutron data library officially distributed by OECD/NEA databank. The approach is oriented towards meeting the broadly accepted general requirements to establish subcriticality, such as those formulated in the ANSI/ANS-8.1-1998 and ANSI/ANS-8.17-2004 Standards.     

Implementation of burn-up credit approach for transport and storage cask

Abstract

TN International currently uses burn-up credit methodology for the design of casks dedicated to the transport of pressurised water reactor uranium oxide spent fuel assemblies. As long as the fuel enrichment of the pressurised water reactor fuel assemblies was sufficiently low, a burn-up credit methodology based on the sole consideration of actinides and the use of a partial burn-up was satisfactory to cover the needs without necessity to design new casks. Nevertheless, the continuous increase in the fuel enrichment during the last decade has led TN International to continue the investigations on the burn-up credit methodology to limit both the increase in the neutron poison content in the new basket designs and the burn-up constraints attached to the acceptability of the fuel assemblies for transport. The strategy of TN International was then to take benefit of the large negative reactivity reserves, which might be gained by the consideration of the fission products coming from the fuel irradiation. A big step forward has recently been reached by TN International on this field with the definition of an advanced burn-up credit methodology based on the consideration of relevant fission products recommended by OECD. In the meantime, TN International has taken the opportunity to use such burn-up credit approach in the design of the TN 24 E transport and storage cask developed for the German nuclear power plants. The relevant task has been carried out according to the German standard DIN 25712 for burn-up credit application. The present paper will describe the basic principles of the burn-up credit methodology implemented by TN International such as:

(i) the current state of the art concerning the burn-up credit application in the criticality assessment

(ii) the basic approach used for the implementation of the advanced burn-up credit methodology (bounding axial burn-up profiles, fuel irradiation parameters, fission products, etc.)

(iii) the area of validity of the TN International burn-up credit approach with fission products

(iv) example of application of the burn-up credit methodology for the design of the TN 24 E transport and storage cask under licensing in Germany

(v) the perspectives of development of the burn-up credit methodology.     

Development of new transportable storage casks for interim dry storage

Abstract

We have started a programme to design a new type of transportable storage cask (Hitz casks) for both boiling water reactor (BWR) and pressurised water reactor (PWR) fuels for use in the new interim dry spent fuel storage project in Japan. The basic policy of this development is to use proven technology to realize a safe and cost-effective design with a high transport and storage capacity and a low fabrication cost. Since it is not permissible to change the lid gaskets at the storage facility, the double-lid system is designed to be able to use double metallic gaskets as the containment boundary for transport after the storage period; this is one of the new design features used in the casks. With the basket design we tried to achieve a capacity of 69 fuel assemblies for BWR fuel and 26 fuel assemblies for PWR fuel. Further details about these and other topics are discussed.     

Operational experience of on-site transport of spent nuclear fuel in Korea

Abstract

With the rapid development of the nuclear power programme in Korea, the amount of accumulated spent nuclear fuel has inevitably increased year by year. The spent nuclear fuel is being stored in on-site storage pools at the nuclear power plants. As the current storage capacity for spent nuclear fuel is insufficient, at-reactor storage is being expanded at each site with regard to optimisation of technical and economic factors. On-site transport between neighbouring reactors has been necessary to secure sufficient storage capacity for pressurised water reactor spent nuclear fuel assemblies. A complete on-site transport system has been developed, and so far more than 800 spent nuclear fuel assemblies have been transported using two kinds of transport cask.     

Transport Calculations of MOX and UO2 Pin Cells by the Method of Characteristics

A new transport theory code for two-dimensional calculations of both square and hexagonal fuel lattices by the method of characteristics has been developed. The ray tracing procedure is based on the macroband method, which permits more accurate spatial integration in comparison to the equidistant method of tracing. The neutron source within each region is approximated by a linear function and linearly anisotropic scattering can be optionally accounted for. Efficient new techniques for both azimuthal and polar integration are presented. The spatial discretization problem in case of P ₁-scattering has been studied. Detailed analyses show that the P ¹-scattering in case of regular infinite array of fuel cells is significant, especially for MOX fuel, while the transport correction is inadequate in case of real geometry multi-group calculations. Finally, the complicated nature of the angular flux in MOX and UO₂ fuel cells is demonstrated.     

New Fresh Mox Fuel Packages: From Design to Operation

Abstract

The MX packages developed by COGEMA Logistics according to TS-R-1 requirements will replace older Current packaging to transport fresh pressurised water reactor (PWR) and boiling water reactor (BWR) mixed oxide (MOX) fuels in Europe. Two types of package have been developed: (1) MX8 with a capacity of eight 17 × 17 900 MWe PWR fuel assemblies for dry loading and underwater unloading operations; and (2) MX6 for dry loading and unloading operations. The capacity of the MX6 is six 16 × 16 or 18 × 18 PWR fuel assemblies or sixteen 10 × 10 BWR fuel assemblies. To meet these capacities requirements, an innovative and optimised design has led to ‘mid-weight’ packages with original solutions for the body, the baskets and the fuel restraining system. To cope with both capacity and legal weight transport requirements, a new high-security transport system has been developed simultaneously. The first shipment with MX8 was made in December 2001, and the first use of MX6 packages is scheduled for the end of 2003.     

Expansion of the capabilities of the GA-4 legal weight truck spent fuel shipping cask

Abstract

General Atomics has developed the model GA-4 legal weight truck spent fuel cask, a high-capacity cask for the transport of four pressurised water reactor (PWR) spent fuel assemblies, and obtained a certificate of compliance (CoC, No. 9226) in 1998 from the US Nuclear Regulatory Commission (NRC). The currently authorised contents for this CoC, however, are much more limiting than the actual capability of the GA-4 cask to transport spent PWR fuel assemblies. The purpose of this paper is to show how the authorised contents can be significantly expanded by additional analyses without any changes to the physical design of the package. Using burn-up credit as outlined in US NRC Interim Staff Guidance 8, Revision 2, the authorised contents can be significantly expanded by increasing the maximum enrichment as the burn-up increases. Use of burn-up credit eliminates most of the criticality imposed limits on authorised package contents, but shielding still limits the use of the cask for higher burn-up, short-cooled fuel. By reducing the number of assemblies transported (downloading) to two and using shielding inserts, even high-burn-up fuel with reasonable cooling times can be transported.     

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.     

PATRAM 2004 — Berlin,September 2004

Abstract

Since British Nuclear Fuels plc (BNFL) was formed in 1971 its transport service has safely moved spent light water reactor (LWR) fuel from many locations abroad to its fuel handling plants at Sellafield in the UK. To support this business a number of types of flasks have been designed and used. One of the types used has been the Excellox family of water-filled flasks. To support future business opportunities a new flask, designed to meet the requirements of the new IAEA transport regulations TS-R-1 (ST-1, Revised), has been developed. The flask will be a type B(U)F. This new flask design will maximise fuel carrying capacity to minimise transport costs. The design capacity of the new Excellox 8 flask is to be 12 pressurised water reactor (PWR) or 32 boiling water reactor (BWR) fuel assemblies. The objective of this BNFL project is to provide another economic spent nuclear fuel transport system, in support of BNFL transport business.