Pre-Normative Research, a Possible Tool for Code Improvement

Wolfgang Hoffelner

(RWH/ Switzerland)

Nuclear Codes and Standards Workshop Prague July 7-8, 2014

Rev. 1 presented by Claude Faidy

Background• increasing demand on components and complexity of operation

(safety concepts, operation conditions, extended lifetime etc.)

• advent of powerful computing capacity

demands on design procedures increased and are still further increasing

• particularly true for nuclear power plants where :– life-time extensions, residual life assessments, damage monitoring

– but also fast neutron spectra, cooling media and higher operation temperatures

provide ambitious engineering challenges which need not only strong support from mechanics but particularly from materials science

• examples for this development :– tasks sponsored by USDOE and NRC in support of Generation IV reactors

– as well as European code related research projects (e.g. MATTER , MATISSE)206.08.2014

Background

• Needs to understand quantatively the behavior of materials during the operation period of a plant are :– not at all limited to one particular code,

– relevant for several design codes and design procedures.

• Code related materials research is an important topic particularly with respect to future code developments.

• Topics like :– system based code, creep-fatigue interactions,

– radiation damage and corrosion,

– assessment and monitoring of damage development

only few examples for design related problems need considerable materials research. 306.08.2014

Some issues for code related materials research

• Unification of allowable stresses

• irradiation damage and corrosion

• Extrapolation of creep and stress rupture data, up to 200 000 hours

• Creep-fatigue interactions

• Subcritical crack growth (creep, fatigue, creep-fatigue)

• Probabilistic aspects of materials damage

• Materials modeling as tool for improved understanding of damage

• Improve collaboration between fission, fusion and thermal plants

406.08.2014

Strength values for the same materials can differ in different codes

Comparison of yield strength values of a carbon steel given in different

standards (GB 713…Chinese Standard, EN 10028-2…European

Standard)

506.08.2014

For low stresses signs for changing creep mechanisms (dislocation creep to diffusion creep) exist which becomes obvious from the deviations from the straight line. Stress rupture data extrapolations can become non-conservative

606.08.2014

Creep-fatigue in components occurs with long hold-times in comparison with typical creep-fatigue experiments. Different life-time prediction

methods lead to very different results and life-time assessments for realistic conditions become therefore difficult

706.08.2014

Consideration of irradiation damagein RCC-MRx

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Negligible Irradiation

Negligible creep

Negligible Irradiation

Significant creep

Significant Irradiation

Negligible creep

Significant Irradiation

Significant creep

Negligible creep

curve

Negligible

irradiation

curve

Maximum

irradiation

curve

Consideration of irradiation damagein RCC-MRx

Negligible Creep Significant creep

Negligibleirradiation

Classical rules (type P damage, type S damage) + notch effect (Fracture mechanics)buckling

type P damage:Sm including correction for thermal ageing S, St : tabulated values = f( θ, t ) type S damage: deformation criteria, fatigue criteria

Significantirradiation

New rules: (type Pdamage, type Sdamage)+notch effect(Fracture mechanics)P+Q et P+Q+F

Extended rules (type Pdamage, type S damage)

New rules (limited domain:material, temperature range)

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Irradiation

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draft RCC-MRx 2010 : Section II - TOME 1 - VOLUME Z

Appendix A3.1S : hypertempered X2CrNiMo17-12-2(N)

2,75; 20 53; 20

53; 375

24; 550

2,6; 375

2,6; 400

2; 425

2; 400

2,75; 375

2; 550

30; 42524; 425

30; 400

40; 400

40; 375

0

100

200

300

400

500

600

0,1 1 10 100dpa

tem

pera

ture

(°C

)

Maximum allowable irradiation

Significant irradiation

Negligible irradiation

Sm Sem (temperature, irradiation parameter )

Multi-scale simulation of radiation damage on steels (LAS, SS)

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From PERFORM 60 EC project

Materials modeling on different scales for damage assessment of nuclearplants, replotted from Roger Stoller (ORNL)

Materials modeling for analysis of radiation damage

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Possible future link between design, residual life assessment and modeling

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Synergies between fission, fusion and thermal technologies

Topic Fission Fusion Conventional

Ferritic/ martensiticsteels

Grade 91 propertiesTMT

Low activation MARBN etc.

ODS F/M Low activation F/M F/M, Nickel base

SiC/SiC X X XMechanical properties X X X

Advanced metallurgical production (e.g.P/M)

X X X

corrosion Sodium, lead, lead bismuth, helium, molten salt, water, supercritical steam

Pb-17Li Gas, steam,

Materials modeling Radiation/ structural stability/mechanical properties

Radiation/structural stability/mechanical properties

Structural stability/mechanical properties (radiation eventually for space applications)

P/M…powder metallurgy, TMT…thermo-mechanical treatment, ODS…oxide dispersion strengthened, F/M…ferritic-martensitic steels

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Conclusions (1 / 2)

• Codes and standards –traditionally build on technical state-of the art knowledge

–and research for future codes has probably not been a high priority subject.

•This might change due to the increasing complexity in plant design and residual life assessment and globalization of design.

•As the mechanical and materials background in the different codes is more or less the same also code requirements have pretty much the same objectives all over the world.

The design procedures might be different, but the background is not !

•It is therefore a real question for the future :more international collaboration between different codes could be of mutual benefit for all of them.

•Harmonization of mechanical properties, consideration of irradiation damage, multi-scale modeling, creep-fatigue interactions, extrapolation of stress rupture data, subcritical crack growth and system based code requirements were considered in this paper as examples.

•Close collaboration between fission, fusion and traditional power plants can have beneficial effects.

1506.08.2014

Conclusions (2 / 2)

•Future actions from the side of ASME could be:

–Watch international activities in the field of related materials research

–Define research topics and projects which might be of interest for future code development

–Participate in international code related research whenever possible and appropriate

•Improve collaboration between fission, fusion and non-nuclear societies (internal and external)

•Already existing and still to be established international ASME groups as well as the Generation IV society might serve as a vessel for promotion of international code related research for the benefit of several codes

1606.08.2014