[en] Improved fuel utilization is possible in tight-lattice PWR cores currently under development within a cooperation of the Karlsruhe Nuclear Research Center (KfK), the Swiss Paul-Scherrer-Institute (PSI), the Siemens/KWU and the Technical University of Braunschweig (TUBS). The hexagonal tightly-packed cores of a Pressurized Water High Converter Reactor (PWHCR) employ higher mass fluxes than conventional PWRs. PWHCR fuel assemblies (FA) if compared with PWR-FA have distinct fluid flow cross sections and obstructions. Also typical heat flux distributions (local peaks) call for appropriate thermal hydraulic models (average and local heat transfer, mixing, CHF). The scarcity of CHF data available for this kind of problem is obvious. For these reasons a broad framed investigation program on tight lattice two-phase mixing and boiling heat transfer has been organized at KfK in a scientific cooperation with TUBS. Performance characteristics of the recently erected Freon test facility KRISTA and some principal CHF results will be discussed to describe actual aspects and goals of the thermal hydraulic model fluid based program. The parallel efforts to develop scaling laws for a transformation of Freon flow boiling results into water equivalent data will be outlined. Final concern is related to the long term replacement of CFC-12 by Ozone-harmless fluids. (author). 17 refs, 10 figs, 2 tabs

[en] Twelve different test zones, Cores 7 to 18, have been studied to date in the Phase II program of LWCHR physics experiments at the PROTEUS zero-power facility. This paper reviews the test lattice configurations investigated, the types of integral measurements carried out, experimental techniques and accuracies, the transferability of results to LWHCR design, as also typical comparisons with calculations based on standard LWR methods. It has been shown that the experimental data base provided is broad - in terms of both high converter design characteristics represented and the types of integral data measured. Thus, the experimental program covers changes in moderation ratio (lattice geometry) and effective fissile-Pu enrichment - with investigations of neutron balance components, moderator voidage effects, influence of lattice poisoning, relative control rod worths and core heterogeneity effects. The importance of having such a broad data base is illustrated by the trends currently reported for the C/E (calculation/experiment) variation for reaction rate ratios with degree of moderation. (author). 18 refs, 3 figs, 6 tabs

[en] In 1981, after the preliminary design stage for the N4 project had been completed, Framatome initiated a research and development effort aimed at defining new core concepts. A first phase was devoted to analyzing the capabilities offered by undermoderated reactors burning plutonium and by spectral shift reactors using uranium. Subsequently the field was broadened in order to anticipate utilities' medium-term requirements. It was concluded that priority should be given to: cost savings, flexibility in utilization of fissile material, operating versatility. At the beginning of the twenty-first century, two main trends will govern fissile material supply: increased availability of plutonium produced by light water reactors (though not in sufficient quantities to cover demand) and the fact that natural uranium will be available at reasonable prices for many years to come. This led us to define the convertible spectral shift reactor concept as early as 1984. Such a reactor uses both types of fissile material to optimum effect, but is still based on existing PWR fabrication and fuel facilities. At about the same time, following the first assessments, a basic research program to investigate the neutronics, thermal hydraulics and safety of tight lattice plutonium cores was launched by the CEA and EdF (DER), bringing in Framatome. The CEA was also involved in Framatome's work on the convertible spectral shift reactor, since it performed the various feasibility demonstration tests. The study was completed in 1988, with the conclusion that the convertible spectral shift reactor is a feasible proposition. This work will be presented in greater detail at the technical meeting. Now, EdF has initiated a program to define the characteristics of a future plant series, the PWR 2000, with the convertible spectral shift reactor as a possible option for the steam supply system. (author). 11 figs, 11 tabs

[en] Two features of the HCR define the main design areas: the triangular pitch of the fuel bundle and the higher density of control assemblies. The triangular pitch requires a novel spacer design. Several solutions have been identified. A honey comb grid assembled from bent straps was selected for further detailed development. In order to obtain specimens for thermohydraulic tests and manufacturing experiences about 50 spacers were built to present. The higher control assembly (CA) density necessary to compensate reactivity in a harder neutron spectrum requires to join at seven positions seven spiders to one ''super RCC'' and to one CA drive. Due to the fact that part of reactivity will be compensated by CA a safety mechanism is required which prevents any rod ejection failure. Furthermore, an adequate coolant flow path especially for the central fuel assembly at a ''super RCC position'', must be provided. A safety mechanism which unlatches drive and control element if the upward motion is larger than a regular step has been designed. The function was demonstrated. The required flow path from the central element into the plenum can be provided by a novel control rod guide structure. A template encloses the complete assembly of seven control rod spiders. The flow area and guide path are separated such that the coolant flow cannot induce adverse vibrations. This structure is extended almost directly from the fuel rod bundle into the upper plenum. The remaining flow path is sufficient. The guide structure itself is latched at the assembly in order to prevent mismatch between the upper guide and the assembly internal guide tubes. (author). 2 refs, 8 figs, 1 tab

[en] The possible option of a tight lattice core for future PWRs has been considered for the last five years, from 1983 to 1987. EdF, FRAMATOME and CEA conducted a joint R and D programme dealing with undermoderated PWR cores. This R and D programme covered the following domains: core physics; thermalhydraulics, both under operating and accidental transient conditions; fuel behaviour and reactor mechanics. The conclusions are reported in two feasibility reports prepared by Framatome (1988) and CEA (1987) which are summarized in the present presentation. The core physics experiments covered the following parameters: material buckling, spectrum index, reactivity and power distribution effects and assembly configuration parameters. The thermalhydraulics results deal with DNB (critical heat flux) on one hand, and with reflood after LOCA on the other hand. As for mechanical design, the main feasibility studies concerned the upper plenum, which is much overcrowded with guide tubes than in the standard design, and the vibratory behaviour of control/spectral shift rod clusters. At present, an important core physics programme is devoted to the support of plutonium recycling in present PWRs. The EPICURE programme provides useful additional results which could be helpful in further potential HCPWR design activities. (author). Figs and tabs

[en] High conversion in water cooled and moderated reactors combines familiar technology backed by extensive experience with desirable feature of efficient fuel utilization. However, economy considerations result in reprocessing installation capacity placing a lower limit on the size of the nuclear programme within which fuel cycles requiring reprocessing can be considered. Direct fuel cycles will therefore remain the preference for small and medium nuclear systems, as well as for a number of developing countries that will be entering the nuclear energy field in 15-20 years time. Feasibility and gains of combining intermediate conversion core of a pressurized water reactor, i.e. semitight lattice, and correspondingly increased fuel utilization with the convenience of open fuel cycle are investigated. At this stage conceptual calculations are in progress with the aim of determining the range of variation of V_m/V_f ratios at the BOL and EOL of fuel compatible with reactivity requirements and with efficient plutonium burning. (author). 11 refs, 3 figs

[en] The use of tight pitch lattice and MOX fuel is one of the ways to improve fuel utilization in WWERs. The presented report is devoted to an analysis of main neutron-physical and thermohydraulic problems connected with such core design. In order to achieve the needed accuracy of tight pitch lattices burnup calculations it is necessary to provide a high accuracy of reaction rates calculations in the resonance energy range and therefore the cross sections of the used Pu, Am, Cm isotopes and fission products should be reliable. For such calculations the UNIRASOS-2 and SAPFIR codes are used at the Kurchatov Institute. The ''second'' equivalence theorem with specially selected parameters and generalized subgroup approach are used for resonance treatment in these codes. MCU code package calculations have been used for their validation. This code package is based on the Monte-Carlo method with a detailed description of cross section energy dependence in the energy range of resolved resonances. A comparison of MCU, UNIRASOS and SAPFIR calculations with the results of precision calculations, results of benchmark problems on tight lattices burnup solutions and measured data obtained at the PROTEUS critical assembly, has shown that the obtained accuracy of these codes is satisfactory for practical purposes. However, in order to estimate the reliability of void reactivity coefficient calculations under low moderator density, it is necessary to perform special investigations. For two dimensional pin power distribution calculations, the 4-group code PERMAK, in which both diffusion and nodal type balance equations are realized, has been used. (author). 30 refs, 4 figs, 6 tabs

[en] One measure to improve fuel utilization in light water reactors is to increase the conversion ratio in a tight, hexagonal PuO₂/UO₂ mixed oxide fuel pin lattice. The PWHCR (pressurized water high converter reactor) is the Siemens/KWU approach towards this kind of tight lattice reactor, with the main characteristics of the actual concept being zirconium-clad fuel rods and an average moderator-to-fuel volume ratio of 1.2. In a recent study, concerning the nuclear core design for the PWHCR, mainly the questions related to the fuel assembly design, the reactivity control system and fuel management strategies have been addressed. Results of the investigations essentially confirmed the concept of the tight lattice PWR to be technically feasible. (author). 5 refs, 8 figs, 4 tabs

[en] Design studies of a high conversion PWR (HCPWR) plant have been performed in which the nuclear characteristics associated with the transition between a semi-tight MOX core and a loose UO₂ core, power capability of the core with respect to Vm/Vf ratio, mechanical integrity of fuel assemblies as well as spectral shift rods within a guide tube under flow-induced vibration in the reactor's upper plenum region, and safety assessment on some of the typical non-LOCA's were evaluated. Technical feasibility of a HCPWR with a semi-tight latticed core is verified whose associated fuel cycle cost may be reducible to the competitive level with respect to an advanced LWR. Further design studies related to the realization of core flexibility with the Vm/Vf ratio ranging between 1.4 and 2.2 are under progress. (author). 5 refs, 11 figs, 4 tabs

[en] Utilization of Pu^f in terms of a thermal recycling reactor, a HCPWR and a FBR is taken into consideration and the contribution to the saving in the annual natural uranium consumption as well as the related average fuel cycle cost over all nuclear power plants in operation is compared under the condition of limited Pu^f supply. HCPWR's introduced in advance of the commencement of introducing FBR's contribute toward the lowering in the annual natural uranium procurement and facilitate the increase in the rate of annual introducible number of FBR plants. A HCPWR without blanket fuel assemblies becomes competitive in fuel cycle cost to the conventional light water reactors by the sacrifice of conversion ratio by 0.05. A HCPWR provided with a flexible core in Vm/Vf ratio, which is achieved simply by replacing the type of fuel assemblies, is recommendable to meet the economic and strategic demands on Pu^f utilization. (author). 5 refs, 6 figs, 2 tabs, 1 diagram