The new code “RCC-MR” — Rules for design and construction of LMFBR components

In 1978, Commissariat à l'Energie Atomique, Electricité de France, and Novatome decided to undertake a common effort to gather a complete collection of rules to apply for design of LMFBR components. The first issue of this work is now being published by AFCEN as the “RCCM” code. The preparation of the design rules used largely the experience gained in Superphenix components analysis, and the results of the large R&D program performed as a support for the design of this plant or at longer term perspective, coordinated by a scientific advisary council of AFCEN (Association Française pour les règles de Conception et de Construction des matériels des Chaudières Electronucléaires).     

Some conclusions with regard to the safety assessment of cracked components drawn from the research programme “integrity of components” (FKS II) atthe present state

The continuation of the research program “Integrity of Components”, Phase II, mainly deals with further evaluation and assessment of material properties and the application of data from small standard specimens to large scale specimens and components. This includes the use of advanced numerical methods to check the transferability of fracture mechanics parameters with regard to the type of load and degree of multiaxiality on the failure behaviour of fracture mechanics specimens with component-like dimensions. Further points of interest are the relationship between upper shelf toughness and load-bearing capacity, the influence of neutron irradiation on the properties, and the effect of corrosion on cyclic crack growth.     

Radioactivity confinement capability of “PARC” concept for the future reprocessing

Aiming at TRU waste arising reduction and economical competitiveness for the future reprocessing, is proposed an advanced process concept which is named PARC (Partitioning Conundrum Key) process. Enhancement of confinement capability for long-lived nuclides in a simplified Purex process is the primary subject of this R&D project. Technologies for long-lived nuclide recovery are under development, focused on ¹⁴C and ¹²⁹I in head end, ²³⁷Np and ⁹⁹Tc in extraction, and ²⁴¹Am the daughter of ²⁴¹Pu in effluents. Those nuclides focused here are mobile in the environment and highly concerned as potential hazardous among the long-lived nuclides in spent fuels. New functions in PARC process concept are designed to mitigate the environmental impacts of reprocessing wastes and also to improve economy of reprocessing in the future.     

Addendum to “an evaluation of alternative nuclear strategies”

This paper presents an assessment by Battelle-Columbus of the technology associated with several reactor concepts which may be considered “advanced” beyond existing LWR's in terms of improved natural resource utilization. The concepts chosen for evaluation and intercomparison are the HTGR, GCFR, MSBR and LWBR. Numerous conclusions may be reached from the study and some interesting trends can be observed. Of greatest significance is the fact that the strategies associated with alternative reactors/fuel cycles will not produce dramatic decreases in short-term fissile demand. A second major conclusion is that all of these advanced systems are considered capable of meeting applicable environmental requirements. A third conclusion is that there is no apparent technical reason for deletion of development efforts on any of these reactors, providing that a commercial interest, complete with significant commitment, is existent.     

On the Lambropoulos-Luco's δ-“no go” conjecture in variational flux synthesis

The Lambropoulos-Luco's δ-“no go” conjecture and related controversies in the domain of the variational flux synthesis with discontinuous trial functions are completely resolved in terms of the functionals, ^(±)δ⁽ⁿ⁾(x − x_o), in the frame of the new spaces of functionals, which are topological dual spaces of discontinuous complex valued functions spaces occurring with left and right continuities.     

Attachment concept for the “multi module segments” in a pulsed DEMO reactor

The pulsed thermonuclear demonstration reactor (DEMO) features challenging operational conditions such as high neutron fluxes, high temperatures, and significant thermo-mechanical stresses. These conditions do not require only a selection of advanced structural materials, but also the development of reliable means to assemble the in-vessel components together; allowing thermal expansions, disassembly, and maintenance in attractive scenarios. Over the course of DEMO lifetime, the materials are subjected to embrittlement by neutron irradiation, swelling, considerable thermo-mechanical fatigue and creep. Traditional joining methods may be rarely used in the harsh fusion environment to assemble different components. In addition any proposed layout should cope with the limited space available inside the vacuum vessel (VV).The objective of this study is to review the proposed attachment systems (developed within the latest European DEMO Conceptual Study) for the vertical segmentation concept called “multi module segments” (MMS). In order to find some place to house the attachments the blanket is cut respecting the Tritium Breeding Ratio limit for tritium self sufficiency. The conditions, neutronic and thermal, in which the attachments are supposed to operate, are calculated. The effects of pulsed operations have also been taken into account. The design of the attachments with the available structural materials with and without an active cooling system is analysed and a new concept for plug/unplug attachments is also suggested.     

Conceptual design of multi-purpose heat reactor “Nuclear Heat Generator”

Small heat reactors can apply to on site demand such as district heat and air conditioning, industrial process heat, greenhouse, and seawater desalination in urban and rural areas. The purpose of this paper is to design conceptually a multi-purpose reactor named “Nuclear Heat Generator (NHG)” which could be installed in energy consuming area. The reactor of 1MWt output is designed without any needs for fuel exchange and decommissioning on site. This cassette typed reactor vessel with sealing is transported to specified fuel fabrication shop every 3 to 4 years in order to exchange used fuels. Steam generators are involved in the self-pressurized integrated reactor with natural circulation. Generated steam pressure from heating reactor is 0.88 MPa (saturated) which is so less than that of current water reactors. Under low steam pressure it is considerably easy to make design of containment vessel and safety device. For economic competition overcoming scale demerit it will be necessary for the cassette type reactor to optimize its system design for the multi-production effect as well as modular construction and recycling system.     

Structural shock tests of prototype FBR “MONJU” scale models

This paper presents some results of experiments which simulate the structural dynamic response of a LMFBR primary coolant boundary to a hypothetical core disruptive accident (HCDA) based on scale models and high explosives. It was noted that high explosives are no longer a good simulant of the HCDA. However, the main purpose of the program, which included this experiment, is not to experimentally predict the dynamic response of the reactor structure at the HCDA, but to validate computer codes, which describe the pressure wave propagation and damage process in the reactor structures, using data obtained from these model experiments. The experiments were undertaken using many 1/15 scale simple models of the reactor vessels and internal structures, as well as 1/15 and 1/7.5 scale complex models of the interim design of prototype LMFBR ‘MONJU’. Simple model experiments involved a series of shock tests using pentolite to investigate the configuration effects of the vessel restraining section, the dipped-plate effect and the core barrel effect, respectively.     

Core performance of new concept passive-safety reactor “kamado” - safety, burn-up and uranium resource problem -

New concept of a passive-safety reactor “KAMADO” has a negligible possibility of core melting and flexibility of total reactor power. The reactor core of KAMADO consists of fuel elements of graphite blocks, which have UO₂ fuel rods and cooling water holes. These fuel elements are located in a reactor water pool of atmospheric pressure (1 atm) and low temperature (< 60°C). In case of LOCA, decay heat from fuel rods is removed by conduction heat transfer to the reactor water pool. Since the cooling water does not contact a fuel rod directly, core design has much flexibility without considering dry-out limitation and Minimum Critical Power Ratio (MCPR). Additionally an effective use of spent fuel is expected.     

PSB-VVER simulation of Kozloduy NPP “loss of feed water transient”

This paper provides a comparison between the PSB test facility experimental results obtained during the simulation of loss of feed water transient (LOFW) and the calculation results received by INRNE computer model of the same test facility. Integral thermal-hydraulic PSB-VVER test facility located at Electrogorsk Research and Engineering Center on NPPs Safety (EREC) was put in operation in 1998. The structure of the test facility allows experimental studies under steady state, transient and accident conditions.RELAP5/MOD3.2 computer code has been used to simulate the loss of feed water transient in a PSB-VVER model. This model was developed at the Institute for Nuclear Research and Nuclear Energy for simulation of loss of feed water transient.The objective of the experiment “loss of feed water”, which has been performed at PSB-VVER test facility is simulation of Kozloduy NPP LOFW transient. One of the main requirements to the experiment scenario has been to reproduce all main events and phenomena that occurred in Kozloduy NPP during the LOFW transient. Analyzing the PSB-VVER test with a RELAP5/MOD3.2 computer code as a standard problem allows investigating the phenomena included in the VVER code validation matrix as “integral system effects” and ”natural circulation“. For assessment of the RELAP5 capability to predict the “Integral system effect” phenomenon the following RELAP5 quantities are compared with external trends: the primary pressure and the hot and cold leg temperatures. In order to assess the RELAP5 capability to predict the “Natural circulation” phenomenon the hot and cold leg temperatures behavior have been investigated.This report was possible through the participation of leading specialists from Kozloduy NPP and with the support of Argonne National Laboratory (ANL), under the International Nuclear Safety Program (INSP) of the United States Department of Energy.     

“Deterioration” criteria for convective heat transfer in gas flow through non-circular ducts

The reduction in turbulent, convective heat transfer parameters observed in some supercritical data and in experiments with common gases can be due to radial property variation, acceleration, buoyancy or combinations of these phenomena, depending on the conditions of the applications. To date criteria for the onsets of these effects have been developed for vertical circular tubes. This note presents extensions of these criteria to non-circular ducts with constant cross-sections as in the cooling channels of some advanced nuclear reactors.     

Low-energy unstable nuclear beam channel “SLOW” at the RIKEN ring cyclotron

A new type of low-energy radioactive nuclear beam channel “SLOW” has been constructed at the RIKEN ring cyclotron facility, intended not only for the study of emission mechanisms of various low-energy radioactive as well as stable isotope ions from a characterized surface of the primary target, but also for the generation of useful radioactive ion beams for surface-physics studies of the secondary target.

In the commissioning experiment of the SLOW beam channel, the reaction products of a heavy-ion induced nuclear reaction have been observed after surface ionization at a hot tungsten target.     

Core concept of a passive-safety fast reactor “METAL-KAMADO” and reactivity coefficients

New concept of a passive-safety simple fast reactor “METAL-KAMADO” with metallic fuels is presented, which has same concept as a passive-safety thermal reactor “KAMADO”. A fuel element of the “METAL-KAMADO” consists of metallic fuel (U–10%Zr) and cooling holes of He gas flow. These fuel elements are located in a reactor water pool of atmospheric pressure (0.1 MPa) and low temperature (<60 °C). In case of LOF, decay heats of fuel elements are removed by natural heat transfer from surfaces of the fuel elements to the reactor water pool.

Preliminary neutronic calculations of the “METAL-KAMADO” show possibility of high burn-up of more than 120 GWd/t with 10% enriched U–Zr fuel. Reactivity coefficients of the core are also discussed.     

Balancing material selection and inspection requirements in structural design of fast breeder reactors based on “System Based Code” concept

The System Based Code concept proposed by Asada et al. [Asada, Y., Tashimo, M., Ueta, M., 2002a. System Based Code—Principal Concept. Proc. ICONE10, 22730; Asada, Y., Tashimo, M., Ueta, M., 2002b. System Based Code—Basic Structure. Proc. ICONE10, 22731] intends to realize margin exchange, in order to optimize design. This paper presents preliminary calculation of margin exchange between material strength and the accuracy and frequency of inservice inspection (ISI), taking a reactor vessel of a fast breeder reactor, of which dominant failure mode is creep-fatigue, as an example. The original design is a structure of forged rings of 316FR, material with superior creep strength. Alternative designs use either Type 304 stainless steel or welded structure of 316FR plates, leading to increased failure probability compared to the original design. The accuracy and frequency necessary to compensate this increase of failure probability was estimated. Results envisioned margin exchange between material strength and ISI under practical conditions. Sophistication of the procedure to calculate failure probability will ensure the application of the concept of margin exchange to practical design.     

European THERFAT project—thermal fatigue evaluation of piping system “Tee”-connections

Thermal fatigue is a significant long-term degradation mechanism in nuclear power plants (NPP). In particular, as operating plants become older and life time extension activities are initiated. Operators and regulators need screening criteria to exclude risks of thermal fatigue and methods to determine significant fatigue relevance. In general, the common thermal fatigue issues are well understood and controlled by plant instrumentation at fatigue susceptible locations. However, incidents indicate that certain piping system Tee's are susceptible to turbulent temperature mixing effects that cannot be adequately monitored by common thermocouple instrumentations. Therefore, a European project on thermal fatigue evaluation of piping system Tee-connections THERFAT has been initiated. In THERFAT, a collation of existing field experience has been conducted leading to a selection of Tee-connections for further investigations. The load determination covers experimental tests and advanced thermo-hydraulic flow simulations. The integrity evaluation work package comprises stress/fatigue analyses and fracture mechanics assessments supported by targeted verification damage tests. Proposals will be made for improved load thermal fatigue assessment procedures, screening criteria to determine lower non fatigue significant threshold values and for establishing a European methodology on thermal fatigue.

THERFAT participants

No.	Name	Organisation	Country
1	K.-J. Metzner	E.ON	Hannover, GER
2	C. Faidy	EDF	Villeurbanne, F
3	J.-A. Le Duff	Framatome-ANP	Paris la Défense, F
4	O. Braillard	CEA	St Paul Lez Durance, F
5	C. Cueto-Felgueroso	Tecnatom S.A.	Madrid, E
6	I. Varfolomeyev	Fraunhofer-Gesellschaft	Freiburg, GER
7	J. Solin	VTT	Espoo, SF
8	M. Schippers	Framatome ANP	Offenbach, GER
9	L. Stumpfrock	MPA	Stuttgart, GER
10	K.-F. Nilsson	EC-JRC/IAM	Petten, NL
11	Y. Hytoenen	Fortum Nuclear Services	Vantaa, SF
12	J. Seichter	Siempelkamp	Dresden, GER
13	T. Abbas	Cinar	London, GB
14	S. Figedy	VUJE	Trnava, SLK
15	P. Carmena	Endesa Generacion	Madrid, E
16	L. Cizelj	Jozef Stefan Institute	Ljubljana, SLO

Full-size table