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International standardization of goods and services is a familiar concept.
To be feasible, standardization requires that a technology be sufficiently mature to employ designs of well-established quality and safety.
This is precisely the case for today's nuclear reactor designs, which represent the culmination of more than 50 years of development.

The concept of standardized reactor designs looks towards a future in which reactors can be built in any country without the necessity of adaptation to specific national regulations. Certainly such standardization will be crucial if nuclear power is to realize its full potential as a major contributor to the clean-energy needs of tomorrow's world. Standardized designs will also contribute to safety in nuclear construction and operations, especially as reactors are deployed in countries that are just beginning to introduce nuclear power.
Achieving reactor design standardization will require the combined efforts of industry, regulators, policymakers, governments and international institutions.
In this paper, the WNA's CORDEL Group proposes a conceptual three-phase programme introducing a mutual acceptance and eventually internationally valid design approvals for standardized reactor designs.

But such an evolution towards internationally valid design approvals would necessarily occur in a manner consistent with each country's sovereignty over its own regulatory framework. Each country's regulator would remain responsible for a comprehensive licensing and oversight process, with a streamlined design approval simply being one part of it. No aspect of the CORDEL proposal is meant to imply that any national regulatory process would be subordinated or limited by foreign decisions.

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During the past 15 years many works has been done on stainless steel fatigue curves :

• in many cases over 105 cycles the RCC-M mean curve has been found un-conservative, in a similar situation than ASME Section III curve [12],
• the fatigue data other 106 cycles can now be done under strain control in a reliable way (improvements in fatigue test machine)
• a large test program has been done in Japan and partially in USA.
• a limited program has been done in France.

This paper presents a new proposal for fatigue stainless steel mean curve in air (the existing ASME curve C seems to cover environmental effect on real structures) and a comparison with the NRC Regulatory Guide 1.207 [14] proposal for new curves.
In order to transfer these mean curve data to structures, different complementary tests have been developed, some under air conditions, some under environmental effects.

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The French field experience in stainless steel bi-metallic welds (BMW) has shown different degradations like external surface corrosion cracks close to the low alloy steel / stainless steel interface or fabrication defects in different other locations.
In many countries, some degradation has been encountered in different type of bi-metallic welds : stainless steel BMW or Nibased alloy BMW through different degradation mechanisms (corrosions).
The critical crack size in different location of a BMW is a key safety issue.
To-day, there is no flaw evaluation procedure for this type of components in existing operation codes, like ASME XI [7], RSE-M [6] or R6 rule [5].
Consequently a fracture mechanic procedures is under preparation in the French RSE-M operation Code [6] in order to evaluate the critical crack sizes of defects in different area of a bi-metallic weld.
The procedure validation is based on 2 specific experimental projects that have been performed on 6" and 16" bi-metallic welds at room temperature and 300°C.
Detailed residual stress measurements and simulation have been done, in order to check their influence on the critical crack size.

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The paper analyses the Quality Assurance/Quality Control requirements imposed by various regulations concerning actual NPP projects and discusses the status of different codes in front of such provisions.

Taking the example of several countries, the paper starts with a general view on
regulations applicable to conventional pressure equipment, and covers in a second stage nuclear pressure equipment. It identifies for manufacturer and third parties, the right QA/QC questioning, before to begin any new nuclear pressure equipment project.

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The design and construction rules for mechanical components of LWR nuclear islands (RCC-M) constantly evolve to reflect the needs of the industry with the objective to fulfill the regulatory demands. Each year, an addendum is thus prepared by Afcen.
The December 2008 addendum includes in particular new grades for products procurement, evolutions on destructive and non destructive examination provisions, consideration of new editions of standards, improvements of text for an easier application.
For a better consistency with regulatory demands, technical code requirements have been updated (pressure test, overpressure protection, examination requirements or material properties), and in certain cases, provisions have been shifted in non-mandatory appendices established for various regulatory contexts for a better adaptation to applications abroad.
In parallel, the conditions for consistency with the European Pressure Equipment Directive (PED) and the French Nuclear Pressure Equipment regulation (ESPN Order dated December 12, 2005) have been deepened and a comparison work was done in the context of the MDEP (Multinational Design Evaluation Procedure initiative of OECD) with equivalent ASME code provisions.
The paper will present the content of the 2008 addendum of RCC-M, and the present status of these two studies on regulatory conformance and international comparisons, as well as some orientations for further evolutions.

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In February/March 2007, The NRC issued Regulatory Guide "RG1.207" and Argonne National Laboratory issued NUREG/CR-6909 that is now applicable in the US for evaluations of PWR environmental effects in fatigue analyses of new reactor components. In order to assess the conservativeness of the application of this NUREG report, Low Cycle Fatigue (LCF) tests were performed by AREVA NP on austenitic stainless steel specimens in a PWR environment. The selected material exhibits in air environment a fatigue behavior consistent with the ANL reference "air" mean curve, as published in NUREG/CR-6909.

LCF tests in a PWR environment were performed at various strain amplitude levels (± 0.6% or ± 0.3%) for two loading conditions corresponding to a simple or to a complex strain rate history. The simple loading condition is a fully reverse triangle signal (for comparison purposes with tests performed by other laboratories with the same loading conditions) and the complex signal simulates the strain variation for an actual typical PWR thermal transient.
In addition, two various surface finish conditions were tested : polished and ground. This paper presents the comparisons of penalty factors, as observed experimentally, with penalty factors evaluated using ANL formulations (considering the strain integral method for complex loading), and on the other, the comparison of the actual fatigue life of the specimen with the fatigue life predicted through the NUREG report application.

For the two strain amplitudes of ± 0.6% and ± 0.3%, LCF tests results obtained on austenitic stainless steel specimens in PWR environment with triangle waveforms at constant low strain rates give "Fen" penalty factors close to those estimated using the ANL formulation (NUREG/6909). However, for the lower strain amplitude level and a triangle loading signal, the ANL formulation is pessimistic compared to the AREVA NP test results obtained for polished specimens. Finally, it was observed that constant amplitude LCF test results obtained on ground specimens under complex loading simulating an actual sequence of a cold and hot thermal shock exhibits lower combined environmental and surface finish effects when compared to the penalty factors estimated on the basis of the ANL formulations.
It appears that the application of the NUREG/CR-6909 in conjunction with the Fen model proposed by ANL for austenitic stainless steel provides excessive margins, whereas the current ASME approach seems sufficient to cover significant environmental effects for representative loadings and surface finish conditions of reactor components.

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