mechanism of degradation of primary components · mechanism of degradation of primary components...

MECHANISM OF DEGRADATION

OF PRIMARY COMPONENTS Marta Serrano

IAEA Training Workshop on

Assessment of Degradation Mechanisms of Primary Components in Water Cooled Nuclear Reactors: Current Issues and Future Challenges

CIEMAT, 29 September – 2 October 2014

M. Serrano, IAEA TW Madrid 2014

Contents

• Introduction

• Radiation damage

RPV

Internals

Void swelling

Irradiation creep

• Stress corrosion cracking, SCC

• Irradiated stress corrosion cracking, IASCC

• Fatigue

• Thermal embrittlement


Introduction

• Why Long Term Operation programmes, two visions: Sustainability LWR: the ability to maintain safe and economic operation of the

existing fleet of NPPs for a longer-than-initially-licensed lifetime - regulatory

High capacity factors and low operating costs make NPPs some of the most safe and economical power generators available. Even when major plant components must be upgraded to extend operating life, these plants often represent a safe, cost-effective, low-carbon asset - utility

• Materials Aging and Degradation programmes to develop the scientific basis for understanding and predicting long-term environmental degradation behaviour of materials in NPPs. To assess the performance of SSCs essential to safe and sustained NPP

operations.

To define operational limits and aging mitigation approaches for materials in NPP SSCs that are subject to long-term operating conditions

M. Serrano, IAEA TW Madrid 2014 INL/EXT-12-24562

SSCs=systems structures and components

Introduction

• Long term operation ageing management programmes covers: Metallic materials in the reactor pressure vessel, its internals and the

primary system piping;

Concrete materials in key structures such as containment, reactor supports and biological shield and spent fuel pool;

Low and medium voltage cable systems.

• Key technical issues long-term operations aging management [EPRI Portfolio 2015] Environmentally assisted cracking (EAC)

Irradiation assisted stress corrosion cracking (IASCC)

Reactor pressure vessel (RPV) embrittlement

Aging of cast austentic stainless steel (CASS)

Aging of concrete structures

Aging of low and medium voltage cables


Introduction

• Ageing management is one part of life

management


Ageing Inspection

Mitigation

Monitoring

Repair

Introduction: Components/Materials

M. Serrano, IAEA TW Madrid 2014 R.W. Staehle

Introduction

• All of the above materials are potentially susceptible

to one or more degradation modes, depending upon

the combinations of material and service conditions.

• Initial microstructure play an important role


Introduction

• Materials already contain

defects

Point defects are where an atom is

missing or is in an irregular place in

the lattice structure.

• self interstitial atoms,

• interstitial impurity atoms,

• substitutional atoms and

• vacancies

Linear defects: Dislocations

Planar defects: Grain boundary

M. Serrano, IAEA TW Madrid 2014 http://www.ndt-ed.org

Introduction

• Initial microstructure


http://www.sut.ac.th

Introduction

• Apart from general corrosion, fatigue and irradiation embrittlement of the pressure vessel, many degradation phenomena were not considered specifically in the design-basis for the current light water reactor

• Key among the degradation phenomena not considered were those associated with corrosion events that were localized either because of metallurgical, stress, environmental, or geometrical conditions.

• Increase the likelihood of material degradation. Long term operation up to 60 years: Higher dose and fatigue cumulative factors

Increase in power output: Increased coolant flow and may, therefore, potentially increase the susceptibility to irradiation assisted stress corrosion cracking (IASCC) of core components, and to flow-accelerated corrosion (FAC) of carbon steel piping and flow-induced vibration (FIV) of other components

longer fuel cycles and decreased outage times may place a constraint on the extent of in-service inspection

M. Serrano, IAEA TW Madrid 2014 NUREG/CR-6923

Degradation mechanisms


IAEA Symposium on Radiation Damage in

Solids and Reactor Materials 7-11 May 1962 in

Venice, Italy.

Degradation mechanisms

• Some degradation mechanisms solved by

mitigation activities

• Some new degradation mechanisms


<<There are known knowns. These are things we know

that we know.

There are known unknowns. That is to say, there are

things that we know we don't know.

But there are also unknown unknowns. There are things

we don't know we don't know.>>

Donald Rumsfeld and Tim Williams

Degradation mechanism

• "Uniform" Corrosion

General Corrosion

Boric Acid Corrosion

Flow-Accelerated Corrosion and Erosion-

Corrosion

• Localized Corrosion

Crevice Corrosion

Pitting Corrosion

Galvanic Corrosion

Microbiologically-influenced Corrosion (MIC)

Environmentally Assisted Cracking

• Intergranular Stress Corrosion Cracking (IGSCC)

• Transgranular Stress Corrosion Cracking (TGSCC)

• Primary Water Stress Corrosion Cracking

(PWSCC)

• Irradiation Assisted Stress Corrosion Cracking

(IASCC)

• Low-temperature Crack Propagation (LTCP)

• Under-clad Cracking and Clad Disbonding

• Fatigue

• Loss of Fracture Resistance

Irradiation Effects

Neutron Embrittlement

Void Swelling Effects

Thermal Aging

M. Serrano, IAEA TW Madrid 2014 NUREG/CR-6923

NUREG/CR-6923, BNL-NUREG-77111-2006 Expert Panel Report on Proactive Materials

Degradation Assessment, February 2007

Degradation mechanism - Interactions

M. Serrano, IAEA TW Madrid 2014 Jennsen

Degradation mechanism - Classification

• Cracking mechanisms

• Embrittlement Mechanisms

• The Dimensional Stability Mechanisms

M. Serrano, IAEA TW Madrid 2014 R. Lott


• Cracking mechanisms

Stress corrosion cracking, SCC (Stress & Environment)

Irradiation stress corrosion cracking, IASCC (Stress &

Environment)

Fatigue (Transient loading, environment)

• Produce observable cracks

• Most probable in region of stress concentration

R. Lott M. Serrano, IAEA TW Madrid 2014


• The Embrittlement Mechanisms

Irradiation Embrittlement (Dose & Temperature)

Thermal Embrittlement (Time, Temperature & Composition)

• Changes in material properties

Strength (increase)

Ductility (decrease)

Toughness (decrease)

Randy Lott M. Serrano, IAEA TW Madrid 2014


• The Dimensional Stability Mechanisms

Void Swelling (Temperature & Dose)

Irradiation Induced Stress Relaxation/Creep (Dose & Stress)

• Component Distortion

• Modify Stress/Strain Distribution

• Affects SCC, IASCC and Fatigue

• The Wear Mechanism

Difficult to compare or rank wear potential in identified

components

R. Lott M. Serrano, IAEA TW Madrid 2014

Radiation damage

• Neutron irradiation causes displacement damage

• In the case of internals in addition generation of He atoms through nuclear transmutation reactions of thermal neutrons mainly with B-10 and Ni-58.


W.J. Weber

Radiation damage

M. Serrano, IAEA TW Madrid 2014 S. Zinkle

Radiation damage - RPV

Aprox 20 m

The reactor vessel used in the first commercial nuclear

power plant, the Shippingport Atomic Power Station.

Photo from 1956.

Olkiluoto 3 EPR™ reactor

Installation of the reactor pressure vessel

http://en.wikipedia.org/wiki/Shippingport_Atomic_Power_Station


• Microstructure

Solute clusters precipitates

Matrix damage

GB segregation

• Mechanical properties

Hardening

Embrittlement


Correlation between SQR of volume fraction of irradiation induced

solute clusters and Charpy Shift Japanese RPV materials

Recent results show that for

Low-Cu steels the formation

of cluster exists (MnNiSi) on

dislocation loops


• Transition temperature – Surveillance

programmes


Radiation damage

• Radiation effects in internals

M. Serrano, IAEA TW Madrid 2014 C. Pokor

Swelling • Void swelling was not considered to be and important degradation mechanism for RPV

internals in the design, mainly due to the lower temperature and neutron dose in comparison to fast reactor where swelling can be an important life limiting issue.

• Recent evidences point out that local gamma heating could increase the temperature of RPV internals up to 370ºC and that the low dose rate typical for PWR would lead to a higher swelling level


Void swelling of PWR materials plotted as a

function of effective full power years [

NUREG/CR-7027 ]

Range of irradiation temperature and dose for which void

swelling data (in color code) have been reported for PWR core

internals

There is no conclusive evidence that void swelling plays an

important role in IASCC of PWR baffle bolts. NUREG/CR-6897

(2006)

Irradiation creep

• Irradiation creep is important in the integrity and functionality assessment of internals, mainly in a synergy way with void swelling and the stresses generated by swelling.

Creep in OSIRIS irradiation conditions is slightly faster than in

BOR-60, and a large reduction in the incubation threshold is also

noted. This difference can be due either to a spectrum or to a flux

effect

F. Garner


Stress Corrosion Cracking, SCC

• SCC is a complex phenomenon driven by the synergistic interaction of mechanical, electrochemical and metallurgical factors.

• Perhaps the most critical factor concerning SCC is that three preconditions are necessary and must be present simultaneously.

• The elimination or reduction of any one of these three factors below some threshold level can, in principle, prevent SCC.

IAEA NP-T-3.13 . SCC in LWR M. Serrano, IAEA TW Madrid 2014

Stress Corrosion Cracking, SCC

• Stages of stress corrosion cracking

M. Serrano, IAEA TW Madrid 2014 R.W. Staehle,

Irradiated Assisted Stress Corrosion Cracking, IASCC

• IASCC is a unique form of SCC that occurs only in highly irradiated components

• Despite over thirty years of international study, the underlying mechanism of IASCC is still unknown

M. Serrano, IAEA TW Madrid 2014 Seong Sik Hwang



Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation

Guidelines (MRP-227-A). EPRI, Palo Alto, CA: 2011. 1022863.

Fatigue

• Fatigue is defined as the structural deterioration that can occur as the result of repeated stress/strain cycles caused by fluctuating loads and temperatures.

• Low-cycle fatigue is defined as cyclic loads that cause significant plastic strain in the highly stressed regions, where the number of applied cycles is increased to the point where the crack eventually initiates.

• When the cyclic loads are such that significant plastic deformation does not occur in the highly stressed regions, but the loads are of such increased frequency that a fatigue crack eventually initiates, the damage accumulated is said to have been caused by high-cycle fatigue.

• The aging effects of low-cycle fatigue and high-cycle fatigue are additive.

• Fatigue crack initiation and growth resistance are governed by a number of material, structural, and environmental factors such as stress range, loading frequency, surface condition, and presence of deleterious chemical species.

• Cracks typically initiate at local geometric stress concentrations such as notches, surface defects, and structural discontinuities.

• The aging effect is cracking.

• Fatigue analyses are potential TLAA (time-limited aging analyses) for Class 1 and selected non-Class 1 mechanical components – considered on the design step

M. Serrano, IAEA TW Madrid 2014 WCAP-17790-NP

Fatigue

• Thermal fatigue is caused by loads conditions related to thermal mixing in piping tees and cyclic thermal stratification that can occur due to swirl penetration of high temperature (hot or cold leg) reactor coolant into attached non-isolable, normally stagnant, branch lines containing lower temperature water.

• The occurrence of thermal fatigue cracking has not resulted in a pipe break, only leakage. However, the costs associated with evaluation, repair and plant unavailability have been significant

• In PWRs the most affected pipes are usually the feed water lines of the steam generator, the pressurizer surge line, and the injection pipes of the emergency core cooling systems

M. Serrano, IAEA TW Madrid 2014 EPRI MRP-24

Fatigue

• Fatigue data indicate significant effects of LWR environment

M. Serrano, IAEA TW Madrid 2014 O. Chopra

Fatigue-environment

• An effect of the environment on fatigue life has been established Environmental fatigue correction factor, Fen, which is defined as the ratio of life in air

at room temperature, N_RTair, to that in water at the service temperature, N_water.

The reduction in life depends on strain rate, DO level in water, and temperature

• The effect of the environment will be considered for new reactors (NUREG CR 6909) and for current operating plants considering applying for licence renewal (for stainless steel NUREG CR 5704) in the USA


Thermal embrittlement

• RPV steels

• Operating temperatures are less than 0.3 times the absolute melting temperature, creep or other self diffusion controlled embrittlement mechanisms are not of concern.

• Two possible aging embrittlement mechanisms are temper embrittlement and strain aging embrittlement. Strain aging in low alloy steels result from interactions between dislocations

and the interstitial atoms of nitrogen and carbon. • RPV - little nitrogen due to vacuum degassing and aluminum grain refinement.

• the potential for strain aging is very moderate and is an unlikely aging concern for carbon steels.

Temper embrittlement is manifested as an increase in DBTT, due to the change from predominantly cleavage fracture to predominantly intergranular fracture along impurity segregation paths.

• Potential for thermal embrittlement for times up to 40 years considered as low for “Western” RPV steels but cannot be entirely dismissed on the basis of the available data.


Thermal embrittlement Cast SS

• Cast austenitic stainless steel (CASS)

CASS and welds have a duplex microstructure consisting of austenite and ferrite phases.

The presence of ferrite phase provides the welds with increased tensile strength and resistance to hot cracking tendencies,

But is also the primary cause of thermal aging embrittlement due to the precipitation of α’ by spinodal decomposition in the ferrite phase.

Additional embrittlement comes largely from carbides and phase precipitation or growth of existing carbides

• While the initial aging effect is loss of ductility and toughness, unstable crack extension is the eventual aging effect if a crack is present and the local applied stress intensity exceeds the reduced fracture toughness.


Thermal embrittlement Cast SS

• Cast austenitic stainless steel (CASS)


NUREG/CR–4513

Summary

• Degradation mechanisms depends on the

combination of material with operating conditions

• Not all the degradation mechanisms were taken

into account on the design

• New degradation mechanisms could appear for

long term operation

• Predictive measurements are needed


mechanism of degradation of primary components · mechanism of degradation of primary components...

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