05 Nuclear fuels (scientific, technical)

found that in the first case, core volume reduces with increasing fuel loadings per plate but r equ i rement of fuel also increases. In the second and third case, core volume as well as fuel r equ i rement decreases with increasing fuel loadings per plate. However in the second case, core volume reduces more rapidly than in case 3 with increasing fuel loadings per plate. Employing s tandard computer code PARET, steady state thermal hydraulic analysis of all these cores was performed. The the rmal hydraulic analysis reveals tha t cores with higher densi t ies and fixed water channel width are bet ter from thermal hydraulic point of view and have fuel and clad t empera tu res within the acceptable limits. But the core with higher densi t ies and op t imum water channel width is a be t te r choice in terms of core compaction, less e3SU loading and higher neut ron fluxes. Finally, the core was compacted in three steps to exploi t the benefi ts of both types of cores. The s t ra tegy resul ted in 36% reduct ion in the core volume, 50% increase in thermal neu t ron flux for i r radia t ion and isotope product ion and a slight reduct ion in e3SU loading. All this was achieved with acceptable peak clad and peak fuel centre l ine tempera tures .

05•02030 Evaluation of the burst characteristics for axial notches on SG tubings Hwang, S. S. et al. Nuclear Engineering andDesign, 2004, 232, (2), 139 143. Some events of s team genera tor tubes have been repor ted in some nuclear power plants a round the world. Main causes of the leakage are from various types of corrosion in the s team genera tor (SG) tubing. Pr imary water stress corrosion cracking (PWSCC) of s team genera tor tubing have occurred in many tubes in Korean plants, and they were repai red using sleeves or plugs. In order to develop proper repair criteria, i t is necessary to ascer ta in the leak behaviour of the tubings. A high-pressure leak and burst test ing system was manufactured. Var ious types of e lec t ro-discharged-machined (EDM) notches having different lengths were machined on the outs ide d iamete r of tes t tubes to study SG tube behaviour. Leak rate and l igament rupture pressure as well as the burst pressure were measured for the tubes at room tempera ture . Rup tu re pressure of the par t through-wall defect tubes depends on the defect depth and length. Water flow rates after the rupture were independen t of the flaw types; tubes having 20 60 mm long E D M notches showed similar flow rates regardless of the ini t ial defect depth. A fast pressur izat ion rate genera ted a lower burs t pressure than the case of a slow pressurizat ion.

05•02031 Extending the operable time of the Syrian MNSR Albarhoum, M. Annals of Nuclear Energy, 2004, 31, (18), 2203 2209. The possibil i ty of extending the operable t ime of the Syrian MNSR is invest igated through a th ree-d imens iona l deta i led model of the reactor constructed for this purpose. Good agreement between calculated and measured reactor pa ramete r values were obta ined for the reactor before modificat ion. The operable t ime is increased by increasing the ini t ial avai lable excess reactivity. The lat ter is increased by adding Top Beryl l ium Shims in the Shim Tray. The increased init ial excess reactivity is compensa ted for by increasing control rod worth by subst i tut ing the actual cadmium absorber by 5 B10 absorbe r . The shut down margin is also enhanced and safer reactor is obtained.

05•02032 Helios, current coupling collision probability method, applied for solving the NEA C5G7 MOX benchmark Ivanov, B. D. et al. Progress in Nuclear Energy, 2004, 45, (2 4), 119 124. As par t of an effort to tes t the abili ty of current t ranspor t codes to t rea t reactor core problems wi thout spat ial homogenizat ion, the latt ice code H E L I O S was employed to perform crit icali ty calculations. The test consists in seven-group calculat ions of the C5 M O X fuel assembly problem specified by earl ier s tudies and comprises two cases - two and three-d imens iona l geometry. There are four fuel assemblies two with M O X fuel, the other two with UO2 fuel. Each fuel assembly is made up of a 17x17 lat t ice of square fuel-pin cells. Fuel pin composi t ions are specified in the Benchmark Specification, which also provides seven- group t ranspor t -correc ted isotropic scat ter ing cross-sections for UO2, the three M O X enr ichments , the guide tubes, the fission chamber and the moderator . This paper preset is the methodology employed in solving the C5G7 M O X Fuel Assembly Problem using the t ranspor t code HELIOS.

05•02033 Homogenization-free reactor core analysiswith general first collision probabilities method Poveschenko, T. S. Progress in Nuclear Energy, 2004, 45, (2 4), 143 152. The C5G7 M O X Bench-mark for current codes has been proposed as a basis to test the abili ty of current t ranspor t codes to teat reactor core problems wi thout spat ial homogenizat ion. This is a seven-group form of the C5G7 M O X fuel assembly problem specified by earl ier studies. There are four fuel assemblies, two contain UO2 fuel e lements and two contain M O X fuel elements . Seven group cross-sections for different kinds of fuel ( three enr ichment of M O X and UO2), the guide tubes, the

302 Fuel and Energy Abstracts September 2005

fission chambers and modera to r are given. Thus this benchmark is jus t a mathemat ica l test tha t allows tes t ing the accuracy of the neu t ron t ranspor t equat ion solut ion with different methods and codes. In this paper the Genera l First Coll ision Probabi l i t ies Method (GFCPM) is used to analyse the two-dimensional configurat ion of this benchmark . A l inear flux approximat ion is used in the reflector. Different calculat ion schemes in the reflector region have been used. The ou tpu t results, I~ff and the pin powers have been analysed. The convergence of the results has been analysed both as a function of the subdivision scheme of the reflector region and of the number of points in the calculat ion scheme for general first collision probabil i t ies . Compar i son has been carr ied out for Keff and pin powers both with the reference results (external convergence) and with the results of different approximat ions of G F C P M (internal convergence).

05•02034 Hydro-thermal-mechanical analysis of thermal fatigue in a mixing tee Chapuliot , S. et al. Nuclear Engineering andDesign, 2005, 235, (5), 575 596. This paper covers work carr ied out by the C E A to study the mechanisms leading to cracking of piping as a resul t of thermal loading in flow mixing zones. The main goal of the work is to analyse, by calculation, the thermal loading caused by turbulent mixing in tees and to unders tand the mechanism of ini t ia t ion and propagat ion of cracks in such components . This work is suppor ted by IRSN. This thermal fat igue phenomenon is still not fully unders tood. One of the main obstacles to its unders tand ing resides in the mul t i -domain nature of the loading and associated damage, involving three complementa ry scientific disciplines: thermal-hydraul ic field, thermo-mechanica l field and mater ia ls science. This paper describes the approach adopted by the C E A to establ ish na tura l mechanisms ( turbulence, puls ing and instabil i ty) which might be the cause of any substant ia l thermo- mechanical loading in the piping. Al though turbulence may be the cause of the thermal s t r ipping (presence of high-frequency the rmal f luctuations on the inner surface of the component) , it cannot a lone explain the propaga t ion of deep cracks. The main reason is the 'high- pass filter ' effect of convection. The wall cannot be subjected to convect ion-rela ted thermal f luctuat ions and frequencies less than the inverse of the turbulence t ransi t time. A s t ra ightforward frequency- based analysis of the loading, carr ied out as a first stage, made it possible to establish the l imits of the loading created by these high- f requency events. However, turbulence can give rise to flow instabil i ty (such as pulsing) of lower frequency. But this cannot explain everything. The geometry ups t r eam of the tee, par t icular ly the sequence of s traight sections and bends can, in certain cases, damp the pulses or great ly amplify them. The use of sui table thermal- hydraul ic model l ing is discussed in the second par t of this article. The final resul t of the thermo-hydro-mechanica l l ink-up on appl icat ion to the complex 3D geometry of the Civaux uni t 1 case (which includes a mixing tee, bends and s t ra ight sections) enabled the observat ions made in this p lant case to be highl ighted and correlated. One of the original i t ies of this s tudy is to carry out the overal l analysis ( thermal- hydraul ic and thermo-mechanica l ) with a single computer code, the CAST3M code developed by the CEA.

05•02035 Improved approach for obtaining rotational components of seismic motion Li, H.-N. et al. Nuclear Engineering andDesign, 2004, 232, (2), 131 137. The ro ta t ional componen t of seismic s t rong-mot ion is a t t ract ing a t tent ion since it is becoming evident that i t may contr ibute considerably to the overal l response of s t ructures to ea r thquake motions. This paper presents an improved method for calculat ing the t ime histories of tors ional and rocking components of ground mot ion corresponding to a set of three recorded or thogonal t rans la t ional components . The mathemat ica l model is based on a deta i led represen ta t ion of soil impedance and contr ibut ions of body waves. The dependence of the angle of wave incidence on the frequency of wave is properly given in the calculat ion of rota t ional components with considera t ion of crit ical incident angles. Numer ica l results of the torsion and rocking obta ined from a set of three recorded t rans la t ional components are also presented.

05•02036 Inelastic analysis of cylindrical steel containment vessels under internal accident conditions Landesmann, A. and de Miranda Batista, E. Nuclear Engineering and Design, 2005, 235, (5), 541 555. The present paper is concerned with the s t ructural safety assessment of a proposed nuclear steel con ta inment shell dur ing a pos tu la ted loss-of- coolant accident scenario. The s t ructural evaluat ion is per formed using a computa t iona l second-order ref ined plast ic-hinge method, which is capable of accurately predic t ing all possible modes of failure in an efficient and computa t ional ly less expensive way than the general FEM formulat ion. A tangent modulus model and a gradual reduct ion of the inelast ic resis tance surface are used to take into account directly the s t ructural s t rength and stabil i ty performances in the e lement formu-

lation. The implemented numerical method provides more reliable safety margins and maintainability, exhibiting a more uniform structural safety level than the linear elastic analysis. A simplified non-linear heat transfer model, developed for symmetrical cross- sections, is used to determine the steel temperature gradient and to establish a link between the thermo and the mechanical analysis. The load resulting from pressure and temperature thermodynamic calcu- lations, obtained for the accident scenario, are considered in the structural quasi static analysis, so that the structural response can be tracked for the entire duration of the simulated accident.

05•02037 Instrumentation and control system design Saito, K. et al. Nuclear Engineering and Design, 2004, 233, (1 3), 125 133. The instrumentation and control system of the high temperature engineering test reactor consists of the instrumentation, control equipments and safety protection systems. There are not many differences in the instrumentation and control equipments design between the high temperature engineering test reactor (HTTR) and light water reactors except for some features. Various kinds of R&D of reactor instrumentation were performed taking into account the HTTR operational conditions, and a plant dynamic analysis was carried out considering the operational conditions of the HTTR in order to design the control system. These systems are required to have a high reliability in respect to safety. In the rise-to-power test it was confirmed that the instrumentation has a high reliability and the control system has a high stability and reasonable damped characteristics for various disturb- ances.

05•02038 Investigations of the SAD design parameters for optimum experimental performance Domangska, G. et al. Annals of Nuclear Energy, 2004, 31, (18), 2127 2138. A project of fast experimental accelerator-driven system called SAD Subcritical Assembly in Dubna containing MOX fuel and driven by 660 MeV proton beam is described and analysed. It is shown by design calculations that the necessity exists for certain modifications, allowing for better reliability of measurements of system time characteristics. Different solutions such as: cadmium separation of the biological concrete shield, admixtures of B203 to the concrete and certain slowing down of neutrons were analysed. Experiments on a bare spallation targets were conducted and the production of radionuclides in the lead target were measured and compared with calculations.

05•02039 Nuclear design Fujimoto, N. et al. Nuclear Engineering and Design, 2004, 233, (1 3), 23 36. The high-temperature engineering test reactor (HTTR) has been designed for an outlet temperature of 950~'C. That is the highest temperature in the world for a block-type high-temperature gas-cooled reactor (HTGR). The functions of the reactivity control system are determined considering the operational conditions, and the reactivity balance is planned so that the design requirements are fully satisfied. Moreover, the reactivity coefficients are evaluated to confirm the safety characteristics of the reactor. The power distribution in the core was optimized by changing the uranium enrichment to maintain the fuel temperature at less than the limit (1600~'C). Deviation from the optimized distribution due to the burnup of fissile materials was avoided by flattening time-dependent changes in local reactivities. Flattening was achieved by optimizing the specifications of the burnable poisons. The original nuclear design model had to be modified based on the first critical experiments. The Monte Carlo code MVP was also used to predict criticality of the initial core. The predicted excess reactivities are now in good agreement with the experimental results.

05•02040 Overview of HTTR design features Shiozawa, S. et al. Nuclear Engineering and Design, 2004, 233, (1 3), 11 21. The Japan Atomic Energy Research Institute (JAERI) designed and constructed the high temperature engineering test reactor (HTTR) in order to establish and upgrade the technology basis for the high temperature gas-cooled reactor (HTGR) and develop the technology for high temperature heat applications. The HTTR is a helium-cooled and graphite-moderated HTGR with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950c'C. The first criticality was attained on November 10, 1998 and the rated power operation with the reactor outlet coolant temperature of 850~'C was achieved on December 7, 2001. Since 2002, safety demonstration tests simulating anticipated operational occurrences such as decrease of primary coolant flow-rate and reactivity insertion have been carried out to ensure safety during off-normal conditions. From 2005, various irradiation tests for fuels and materials will start. In addition, a hydrogen production test facility will be coupled with the HTTR by

05 Nuclear fuels (scientific, technical)

2015 to produce hydrogen by nuclear. The history and future plan, major design features and R&D programs of the HTTR are summarized in this paper.

05•02041 Parameter estimation during a transient - application to BWR stability Tambouratzis, T. and Antonopoulos-Domis, M. Annals of Nuclear Energy, 2004, 31, (18), 2077 2092. The estimation of system parameters is of obvious practical interest. During transient operation, these parameters are expected to change, whereby the system is rendered time-varying and classical signal processing techniques are not applicable. A novel methodology is proposed here, which combines wavelet multi-resolution analysis and selective wavelet coefficient removal with classical signal processing techniques in order to provide short-term estimates of the system parameters of interest. The use of highly overlapping time-windows further monitors the gradual changes in system parameter values. The potential of the proposed methodology is demonstrated with numerical experiments for the problem of stability evaluation of boiling water reactors during a transient.

05•02042 Performance test of HTTR Nakagawa, S. et al. Nuclear Engineering and Design, 2004, 233, (1 3), 291 300. The high temperature gas-cooled reactor (HTGR) is particularly attractive due to its capability of producing high temperature helium gas and due to its inherent safety characteristics. The high temperature engineering test reactor (HTTR), which is the first HTGR in Japan, was successfully constructed at the Oarai Research Establishment of the Japan Atomic Energy Research Institute. The HTTR achieved full power of 30 MW at a reactor outlet coolant temperature of about 850c'C on 7 December 2001 during the 'rise-to-power tests'. Two kinds of tests were carried out during the 'rise-to-power tests'. One is commissioning test to get operation permit by the government and another is test to confirm a performance of the reactor, heat exchanger, and control system. From the test results of the 'rise-to-power tests' up to 30 MW, the functionality of the reactor and the cooling system were confirmed, and it was also confirmed that an operation of reactor facility could be performed safely.

05•02043 Potential of thorium molten salt reactorsdetailed calculations and concept evolution with a view to large scale energy production Nuttin, A. eta[. Progress in Nuclear Energy, 2005, 46, (1), 77 99. The paper discusses the concept of a thorium molten salt reactor dedicated to future nuclear energy production. The fuel of such reactors being liquid, it can be easily reprocessed to overcome neutronic limits. In the late sixties, the MSBR project showed that breeding is possible with thorium in a thermal spectrum, provided that an efficient pyrochemical reprocessing is added. With tools developed around the Monte Carlo MCNP code, the performance of a MSBR-like

3 3 reference system with 2 2Th / 2 2 U fuel was re-evaluated. An important reduction of inventories and induced radiotoxicities was found at equilibrium compared to other fuel cycles, with a doubling time of about 30 years. The study then considered how to start this interesting reference system with the plutonium from PWR spent fuel. Such a transition appears slow and difficult, since it is very sensitive to the fissile quality of the plutonium used. Deployment scenarios of Z3ZTh/ 232U MSBR-like systems from the existing French PWRs demonstrate the advantage of an upstream 232U production in other reactors, allowing a direct start of the MSBR-like systems with 232U. This finally leads to the exploration of alternatives to some MSBR features, for energy production with 232Th/ 232U fuel from the start. Different options were then tested, especially in terms of core neutronics optimization and reprocessing unit adaptation.

05•02044 Purex co-processing of spent LWR fuels: flow sheet Zabuno~lu, O. H. and Ozdemir, L. Annals of Nuclear Energy, 2005, 32, (2), 151 162. Purex co-processing of spent LWR fuel is investigated. In purex co- processing, uranium and plutonium in spent fuel are processed and recovered together as a single stream, while in standard purex reprocessing uranium and plutonium are obtained as separate streams. A two-step (co-decontamination and co-stripping) flow sheet for purex co-processing is devised; concentrations, recoveries and decontamina- tion factors are calculated; and methods to co-convert uranium plutonium nitrate to mixed oxide are reviewed. A closed nuclear fuel cycle in which at no point uranium and plutonium are separated from each other is reached.

05•02045 R&D on core seismic design Iyoku, T. et al. Nuclear Engineering and Design, 2004, 233, (1 3), 225 234.

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