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).     

A study of fuel behavior in PWR design basis accident: an analysis of results from the PHEBUS and EDGAR experiments

As part of the French PWR safety study programme, fuel behavior during a design basis accident has been investigated in three parallel directions:

• - separate effect tests in the EDGAR apparatus for developing and validating cladding deformation models,

• - integral tests in PHEBUS for verifying codes,

• - development of fuel behaviour codes for plant calculation after assessment against experimental results. After describing the objectives and content of each of these programmes, the main findings are highlighted and discussed.

Hybrid soliton nuclear reactors: A model and simulation (encapsulated long living accelerator driven system)

We intend to explore the potential of Hybrid Soliton Reactors (Réacteur Hybride à Soliton, RHYS) for producing energy. In our case an encapsulated long living fission reactor is driven by a proton accelerator, who produces neutrons on a target. In a first part we give the mathematical approach of such a sub-critical reactor, as an extension of the “Soliton Reactor” which was recently proposed by different authors, Edward Teller, L.P. Feoktistov, and others (H. Sekimoto under the name “Candle reactor”). In a second part we give results of simulations and explore the possibilities to control such a system.     

Evaluation of ASME code flaw analysis procedure using the influence function method for application to PWR primary piping

This paper discusses stress intensity factor calculations and fatigue analysis for a PWR primary coolant piping system. The influence function method is applied to evaluate ASME Code Section XI Appendix A “analysis of flaw indication” for the application to a PWR primary piping. Results of the analysis are discussed in detail.     

RCC-M (rules for design and construction of nuclear components)

The paper deals with a presentation of the design rules included in the French RCC-M code applicable to mechanical components of PWR nuclear islands and published by the French Society for Design and Construction rules for Nuclear Island Components (AFCEN). Particular attention is paid to the major principles which constitute the background of the rules of the code and to recent developments included in the code.     

SCALE calculation for evaluation of spent fuel condition during long-term storage

Experiences with an advanced spent nuclear fuel management in Slovakia are presented in this paper. The evaluation and monitoring procedures are based on practices at the Slovak wet interim spent fuel storage facility in NPP Jaslovské Bohunice. Since 1999, leak testing of WWER-440 fuel assemblies were completed using a special leak tightness detection system developed by Framatome-anp, “Sipping in Pool”. This system utilized external heating for the precise defects determination.Optimal methods for spent fuel disposal and monitoring were designed. A new conservative factor for specifying of spent fuel leak tightness is introduced in the paper. Limit values of leak tightness were established from the combination of SCALE4.4a (ORIGEN-ARP) calculations and measurements from the “Sipping in Pool” system. These limit values are: limiting fuel cladding leak tightness coefficient for tight fuel assembly – k_FCT(T) = 3 × 10⁻¹⁰, limiting fuel cladding leak tightness coefficient for fuel assembly with leakage – k_FCT(L) = 8 × 10⁻⁷.     

The use of the uniform risk spectra in the seismic PSA for a PWR power station

This paper describes a simple method for incorporating the effects of the uniform risk spectra (URS) in the seismic probabilistic safety assessment (PSA) for a pressurized water reactor (PWR) power station. The “traditional” fragility parameters for a range of critical equipment items in a PWR power station on two typical UK sites are modified to incorporate the URS using this simple method and the effect on the high confidence low probability of failure (HCLPF) acceleration levels and seismic-induced failure probabilities of the equipment items is examined. The results illustrate the potential benefit of using the URS in the seismic PSA for a PWR power station.     

Iodine behaviour under LWR accident conditions: Lessons learnt from analyses of the first two Phebus FP tests

The International Phebus Fission Product programme, initiated in 1988 and performed by the French “Institut de Radioprotection et de Sûreté Nucléaire” (IRSN), investigates through a series of in-pile integral experiments, key phenomena involved in light water reactor (LWR) severe accidents. The tests cover fuel rod degradation and the behaviour of fission products released via the primary coolant circuit into the containment building.The results of the first two tests, called FPT0 and FPT1, carried out under low pressure, in a steam rich atmosphere and using fresh fuel for FPT0 and fuel burned in a reactor at 23 GWdt⁻¹ for FPT1, were immensely challenging, especially with regard to the iodine radiochemistry. Some of the most important observed phenomena with regard to the chemistry of iodine were indeed neither predicted nor pre-calculated, which clearly shows the interest and the need for carrying out integral experiments to study the complex phenomena governing fission product behaviour in a PWR in accident conditions. The three most unexpected results in the iodine behaviour related to early detection during fuel degradation of a weak but significant fraction of volatile iodine in the containment, the key role played by silver rapidly binding iodine to form insoluble AgI in the containment sump and the importance of painted surfaces in the containment atmosphere for the formation of a large quantity of volatile organic iodides.To support the Phebus test interpretation small-scale analytical experiments and computer code analyses were carried out. The former, helping towards a better understanding of overall iodine behaviour, were used to develop or improve models while the latter mainly aimed at identifying relevant key phenomena and at modelling weaknesses. Specific efforts were devoted to exploring the potential origins of the early-detected volatile iodine in the containment building. If a clear explanation has not yet been found, the non-equilibrium chemical processes favoured in the primary coolant circuit and the early radiolytic oxidation of iodides in the condensed water films are at present the most likely explanations. Models that were modified or developed and embodied in the computer codes for organic iodide formation/destruction in the gas phase and Ag–I reactions in the sump lead, in agreement with the Phebus findings respectively to greatly enhanced organic iodide formation kinetics and long term concentration in the containment atmosphere on one hand and, in the conditions of Phebus experiments, to significantly limited molecular iodine volatilisation from the sump in so far as silver was in excess compared to iodine, on the other hand. Organic iodides then quickly gain in importance and become the predominant volatile iodine species at long term.     

An in-pile sodium loop for investigation of local cooling disturbances in LMFBR's

In the safety analysis of Liquid Metal Fast Breeder Reactors, investigations of the fuel element behavior under local off-normal cooling conditions and the possible failure propagation are of special interest. In a common program, called “Mol 7C” the Gesellschaft für Kernforschung, Karlsruhe, and the Centre d'Etude de l'Energie Nucléaire/Studiecentrum voor Kernenergie, Mol, are performing related in-pile experiments in a sodium loop in the BR 2-reactor. The test section contains a 37-rod bundle of fresh UO₂-fuel. A local blockage within the fuel bundle will initiate a certain local damage to a few rods. The experiments are expected to obtain important informations with respect to the problems of pin to pin propagation and the long term behaviour of a fuel bundle with defect pins. The in-pile part of the loop contains the fully integrated primary sodium circuit. Total heat removal capacity is about 700 kW. The equipment for the first experiment is nearly manufactured. The first experiment will start in the beginning of 1977. At first three experiments are planned.     

Standardized French PWR programs and the application of mechanical analysis

The implementation of the French PWR construction programs is marked by the following factors:

• - the construction of similar plants by series, together with the standardization of plant layout and component design,

• - the various aspects of the plant operation policy adopted by Electricité de France (EdF), the main client,

• - the special provisions of French licensing regulations,

• - the continuous development of technical experience required to support an active export policy.

The effect of these factors on the organization of the programs and on technical achievements will be examined, with special attention being given to the mechanical aspects, which constitute the main subject of the SMiRT conference.     

Neutron noise measurements on pressurized water reactors

Noise analysis contributes to increase significantly understanding of safety and monitoring of PWR. The difficulties of a correct interpretation of noise signal in a power reactor encourage a deeper insight into the theoretical model. Following, this paper is dealing with 4 topics:

• - theoretical model

• - measurements made in PWR's, and evolution of power spectral densities,

• - experimental test of models,

• - future study in this range : methodology for early detection of failure and to indicate incipient failure.

Using the expressions for linear system and for feedback loop, we obtain a simplified diagram for PWR with reactivity, temperature, velocity of fluid inputs and movement of internal structures.

Neutron noise measurements are performed periodically on the “Centrale Nucléaire des Ardennes”. An investigation is also performed, in order to detect core barrel movements.

The change of neutron noise at several power levels is shown, and used to check the model.

The development of signal analysis and models for PWR is investigated in the last chapter.     

ATWS analysis for PWR using the coupled code system DYN3D/ATHLET

The ATWS transient “Loss of main feed water supply” in a generic four-loop PWR at the nominal power of 3750 MW was analyzed using the coupled code system DYN3D/ATHLET. A variation of the MOX-fuel-assembly portion in the core has an effect on the reactivity coefficients of the fuel temperature and the moderator density. These two parameters mainly influence the behaviour of the coolant pressure, which is safety-relevant. It has been demonstrated that the pressure maximum decreases with an increasing portion of MOX. For all core loadings considered, both primary-circuit mechanical integrity and sufficient core cooling are guaranteed.     

Development and application of filters for air cleaning in nuclear power plants

The development of filter systems for air cleaning in nuclear power plants will be briefly described. The result of research work on iodine filters was the basis for the use of gasketless deep-bed filters and of the multiway sorption filter in German reactor stations. The composition of the iodine release to the environment was validated with the “discriminating iodine species sampler”. The main sources for the release of elemental iodine from BWR and PWR were discovered and finally suppressed. The mechanical strength of HEPA filters is being tested at high temperatures and humidities. Prototype HEPA filters have been developed with much higher resistance against humidity, differential pressure, and corrosion. A filter for the removal of particles during extreme operating conditions such as containment venting was developed using stainless steel fibers for the filter medium. The first filter of this type has already been built and installed in a modern German PWR as a part of the containment-venting system for serious accidents.     

Critical heat flux and heat transfer above mixture level under high-pressure boil-off conditions in PWR type and tight-lattice type fuel bundles

Heat transfer tests were conducted in PWR 17 × 17 type and tight-lattice type fuel bundles under high-pressure boil-off (very-low flow, mass fluxes lower than 100 kg/m²s) conditions. There is almost no significant difference in both critical heat flux (CHF) (or dryout point) data and convective heat transfer data above the mixture level between the PWR type and tight-lattice type bundles. The “complete vaporization equation” predicts well the CHF data, i.e. the dryout occurs nearly at the elevation where the thermal-equilibrium quality reaches 1.0. The Groeneveld CHF table used in the RELAP5/MOD3 code should be improved in the region of mass flux between 10 and 100 kg/m²s. The radiative heat transfer has an important contribution to total heat transfer above the mixture level. The Dittus-Boelter correlation, with use of the film temperature in evaluating steam properties, predicts well the convective heat transfer above the mixture level.     

Advanced three-dimensional two-phase flow simulation tools for application to reactor safety (ASTAR)

This paper summarizes the cumulative work undertaken in the frame of the EU shared-cost action “ASTAR Project”—the current status and future perspectives in the field of advanced numerical simulation of three-dimensional two-phase flow processes. This 3-year running project, which started in September 2000, involves seven partner institutes from around Europe. Specific emphasis is given to the further development of characteristic-based upwind differencing (also called “hyperbolic”) numerical methods and their application to transient two-phase flow. The paper summarizes the common basis adopted for the physical and mathematical modelling of two-phase flow in the form of a single-pressure “two-fluid” model and the various numerical solution techniques developed by the partners. Several benchmark exercises are presented which have been used as verification and assessment procedures for comparing the different modelling and numerical approaches. Comments on the suitability, accuracy, numerical stability, algorithmic robustness and computational efficiency serve as indicators for the possible extension of these methods to future code development activities. Two further tasks of the ASTAR project dealt with the production of high quality experimental field data in the LINX facility of PSI, for the validation of CFD models for two-phase bubbly flow, and the coupling of a two-phase CFD module with a system code. Details of these tasks have been published separately, and will not be recalled in this paper.     

Fission product release under severe accidental conditions: general presentation of the program and synthesis of VERCORS 1–6 results

The French Nuclear Protection and Safety Institute (IPSN) launched the HEVA-VERCORS program in 1983, in collaboration with Electricité de France (EDF). This program is devoted to the source term of fission products (FP) released from PWR fuel samples during a sequence representative of a severe accident. The analytical experiments are conducted in a shielded hot cell of the LAMA facility of the Grenoble center of CEA (Commissariat à l’Energie Atomique); as simplified tests addressing a limited number of phenomena, they give results complementary to those of the more global in-pile PHEBUS experiments. Six VERCORS tests have been conducted from 1989–1994 with higher fuel temperatures (up to 2600 K) compared with the earlier HEVA tests in order, in particular, to quantify better the release of lower volatile FPs. This paper gives an overview of the experimental facility, a synthesis of FP release from these tests and exhibits, as an example, some specific results of the VERCORS 6 test, performed with high burn-up fuel (60 GWd tU⁻¹). The on-going VERCORS HT–RT program, designed to reach fuel liquefaction temperatures, is described before conclusions are drawn.     

UPTF experiment refined PWR LOCA thermal-hydraulic scenarios: conclusions from a full-scale experimental program

A total of 34 tests were performed at upper plenum test facility (UPTF, a 1:1 scale test facility) to investigate the thermal-hydraulic phenomena in a pressurized water reactor (PWR) primary system during end-of-blowdown, refill and reflood phases of a loss-of-coolant accident (LOCA). Separate effect tests as well as integral tests were carried out. After the completion of the program a summary of the basic findings from the full-scale tests is given, focusing on thermal-hydraulic issues related to: two-phase flow phenomena at the ECC injection ports for cold or hot leg injection; the ECC delivery into core area via the downcomer or the tie plate; entrainment de-entrainment phenomena during reflood (i.e. the “steam binding” and driving water head reduction problems).     

On-line monitoring the incore power distribution by using excore ion-chambers

For most of nowaday pressurized water reactors (PWR), excore ion-chambers (EIC) are the sole real-time sensors responding to the incore power distribution (IPD). Conventionally, it is supposed that the EIC could not carry out information of the IPD in radial direction. As a result, there is an excess conservatism when they are being used for the core monitoring so far, and some flexibilities of the reactor operation are sacrificed. Though progresses have been made, there are still some inevitable obstacles about newly developed monitoring methods: mounting and maintaining a large amount of fixed incore detectors are costly; and the IPD calculated by on-line simulators is not the “measured” one. Fortunately variations in the IPD must be driven by reasonable physical causes, and the corresponding variations in the readings of the EIC are notable enough to be identified. So with a specific referenced IPD measured from periodical flux mappings, it is possible to on-line monitor such variations by using the EIC. Aimed at this, a harmonics grouping method and an algorithm of higher order perturbation variables are developed, and an influence equation for control rods is also proposed. Then a precise restriction equation describing the reasonable variations in the IPD is deduced for example of the Heating Reactor. Combined with a proper-configured EIC, a systematic “excore-to-incore” method is realized finally.     

Experimental studies of PWR primary piping under loca conditions

The experimental program performed on AQUITAINE-II facility is directed to study the mechanical behavior of primary PWR pipes and the forces exerted on the neighbouring structures as a consequence of a breach opening. It is jointly financed by the Commissariat à l'Energie Atomique, Framatome, Electricité de France and Westinghouse. Some forty tests have been carried out with different pipe configurations (straight tube, elbow, S-or U-shaped tube) and different break types (single or double guillotine). The following aspects are investigated: (1) the dynamic behavior of the pipe and in particular the formation of a plastic hinge at the restraint; (2) impact function of a pipe on an energy-absorbing bumper; (3) lateral stability of both ends of a pipe, after a double-guillotine break.