Materials and Processes for Nuclear Energy Today and in the Future E-Book

• Herausgeber: John Wiley & Sons
• Kategorie: Wissenschaft und neue Technologien
• Serie: ISTE Consignment
• Sprache: Englisch

Seitenzahl: 550

Veröffentlichungsjahr: 2024

Beschreibung

As a low carbon energy source, nuclear energy plays a reinforced role in a sustainable electricity mix. However, strengthening the share of nuclear energy implies the guarantee of safe, long-term operation of current systems and potentially the fostering of new constructions. Service life extension – as well as the design of future nuclear power plants – relies on the availability of robust and qualified structural materials, and their manufacturing processes. The science and engineering of materials are key in selecting robust material solutions and predicting aging mechanisms.

Materials and Processes for Nuclear Energy Today and in the Future reviews different reactor concepts and fuel management systems. Nuclear equipment has to maintain integrity under extreme conditions, such as high temperature, radiation, loads and/or corrosive environments. This book analyzes the requirements on components, and introduces reference solutions regarding materials and processes. It describes the materials’ main properties, their limits and the current R&D trends. Lastly, innovations are discussed, such as materials with enhanced properties, advanced manufacturing or using AI.

Leseprobe

SCIENCES

Energy, Field Directors – Alain Dollet and Pascal Brault

Raw Materials and Materials for Energy,Subject Head – Frédéric Schuster

Materials and Processes for Nuclear Energy Today and in the Future

Coordinated by

Fanny Balbaud-Célérier

Céline Cabet

First published 2024 in Great Britain and the United States by ISTE Ltd and John Wiley & Sons, Inc.

Apart from any fair dealing for the purposes of research or private study, or criticism or review, as permitted under the Copyright, Designs and Patents Act 1988, this publication may only be reproduced, stored or transmitted, in any form or by any means, with the prior permission in writing of the publishers, or in the case of reprographic reproduction in accordance with the terms and licenses issued by the CLA. Enquiries concerning reproduction outside these terms should be sent to the publishers at the undermentioned address:

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© ISTE Ltd 2024The rights of Fanny Balbaud-Célérier and Céline Cabet to be identified as the authors of this work have been asserted by them in accordance with the Copyright, Designs and Patents Act 1988.

Any opinions, findings, and conclusions or recommendations expressed in this material are those of the author(s), contributor(s) or editor(s) and do not necessarily reflect the views of ISTE Group.

Library of Congress Control Number: 2024932616

British Library Cataloguing-in-Publication DataA CIP record for this book is available from the British LibraryISBN 978-1-78945-186-3

ERC code:PE8 Products and Processes Engineering PE8_6 Energy processes engineeringPE4 Physical and Analytical Chemical Sciences PE4_14 Radiation and Nuclear chemistry

Preface

Fanny BALBAUD-CÉLÉRIER1 and Céline CABET2

1 CEA, Service de recherche en Corrosion et Comportement des Matériaux, Université Paris-Saclay, Gif-sur-Yvette, France

2 CEA, Service de Recherche en Matériaux et Procédés Avancés, Université Paris-Saclay, Gif-sur-Yvette, France

In the actual context of climate change, with the urgent need for low carbon energy technologies to fulfill the global clean energy demand, nuclear energy is an important contributor to the energy technology mix. Indeed, the continuous growth of the earth’s population, improvement of living standards and longer life expectancy require the urgent deployment of safe and sustainable low carbon energy technologies. Nuclear power has been part of the global energy mix for more than 50 years; currently it represents 10% of electricity generation, with more than 440 reactors worldwide, and is the world’s second largest source of low carbon electricity. In certain countries, it may thus be an important low emission source of electricity, which can complement and back renewable energies in decarbonizing the power sector.

The mean age of the world nuclear fleet is 31 years old (sources: WNISR, with IAEA-PRIS, 2022), and even though the design life of these power plants is typically 40 years, many nuclear power plants will be able to operate longer. This safe and economic operation is essentially due to the initial choices made about the materials used for their construction. Life extension of functioning nuclear power plants is a reality in many countries and new builds are planned. Moreover, many countries have interest in advanced systems, such as small modular reactors and next generation systems. These small modular systems could allow for the production of low carbon heat and hydrogen.

In this framework, material science and engineering, including manufacturing processes, play a crucial role in safe, reliable, efficient, economical and overall sustainable nuclear energy. The first priority when performing materials selection for nuclear reactors is to ensure continuous safe operation of the nuclear system, considering the extreme conditions these materials encounter, such as high temperature, high irradiation flux and dose, high stress and chemically aggressive environments. Extending service life requires a thorough understanding of material aging mechanisms to ensure safe operation over time. In many cases, the possibility of component inspection is limited and the lifetime has to be certified over decade-long durations. To this end, modeling and simulation of the behavior of non-replaceable components will be essential, together with surveillance programs.

Moreover, new materials solutions are also under study, including advanced materials processing, innovative materials, and functionalized and architectured systems. The miniaturization of samples and testing systems is also under development to facilitate nuclear testing. Searching for new materials and processes, tailoring them to the desired multifunctional properties and processing them into complex-shape components is central to the nuclear sector. Workability, weldability, cost, etc., are other important aspects that need to be looked into during the materials selection process. These material solutions need to be scalable from lab to pilot to demo industrial size, and must also be cost effective. They must also be developed or designed, considering their overall environmental sustainability. Accelerating the development of materials may be a game changer in reducing the time to market of material innovations. Approaches integrating digital materials design using artificial intelligence tools, high-throughput manufacturing processes and accelerated qualification technologies offer real opportunities in the nuclear field over the next few years.

Access to robust and qualified materials, together with their processing, is critical for many major technological challenges the nuclear sector is facing: (i) the long-term operation of existing Generation II+/III nuclear power plants; (ii) the reprocessing, intermediate storage and final disposal of nuclear wastes; (iii) the construction of the next generation of reactors: generation IV power reactors, small modular reactors, advanced modular reactors; and (iv) the design of fusion power plants of tomorrow.

This book describes the structural materials and processes considered for different existing and prospective nuclear systems. The systems are globally introduced and illustrated. The operating conditions and technological constraints set a series of requirements on materials, which are developed and justified. The current choices of materials and their main properties are explained, together with the manufacturing routes and joining techniques. The main limitations of the materials are examined and the chapters conclude with the current R&D trends that are being used to try to overcome these restrictions. The last chapters more specifically contemplate the innovation pathways in terms of novel materials and the accelerated discovery of materials.

Chapter 1 provides an extended review of materials and processes for light water reactors. Chapter 2 includes materials and processes for nuclear fuel reprocessing. Chapter 3 discusses materials and processes for geological disposal of high-level waste and spent fuel. Chapter 4 develops materials and processes for sodium cooled fast reactors. Chapter 5 is dedicated to materials and processes for heavy liquid metal cooled reactors. Chapter 6 deals with materials and processes for molten salt reactors. Chapter 7 addresses materials and processes for fusion reactors. Chapters 8 and 9, respectively, cover advanced materials in terms of ceramics and composite materials, and oxide-dispersed strengthened steels. Chapter 10 concludes with materials discovery, with three parts on materials discovery/numerical design, advanced manufacturing processes and high throughput characterization.

We would like to extend our warmest thanks to all the authors and contributors to this book. Without their participation, expertise and enthusiasm, this work would not have been possible.

April 2024

1Materials and Processes for Light Water Reactors

Jean-Paul MASSOUD1 and Eric MOLINIÉ2

1 EDF/DIPNN/DT, France

2 EDF/DIRES/R&D, France

1.1. Acronyms

BCC:

body centered cubic (cristallography)

BOP:

balance of plant

BWR:

boiling water reactor

CCS:

components cooling system

CRDM:

control rod drive mechanism

CVCS:

chemical and volume control system

CW:

cold worked

DBTT:

ductile brittle transition temperature

DPT:

dye penetrant testing

EAF:

environmentally assisted fatigue

ECT:

eddy current testing

EDF:

Electricité de France

ESW:

electroslag welding

ESWS:

Essential Service Water System

FAC:

flow-assisted corrosion

FCC:

face centered cubic (cristallography)

GTAW:

gas tungsten arc welding

HAZ:

heat-affected zone

HIP:

hot isostatic pressing

IASCC:

irradiated-assisted stress corrosion cracking

IGSCC:

intergranular stress corrosion cracking

LAS:

low alloy steel

LPSIS:

low-pressure safety injection system

LWR:

light water reactor

MPT:

magnetic particle testing

MSR:

moisture separator-reheater

NPP:

nuclear power plant

NSSS:

Nuclear Steam Supply System

PHT:

preliminary heat treatment

PWR:

pressure water reactor

PWSCC:

primary water stress corrosion cracking

PZR:

pressurizer

RBWMS:

Reactor Boron and Makeup Water System

RCS:

reactor coolant system

RHR:

reactor heat removal

RPV:

reactor pressure vessel

RT:

radiographic testing

RT

NDT

:

reference temperature Nihil Ductility Temperature

SAW:

submerged arc welding

SCC:

stress corrosion cracking

SG:

steam generator

SMAW:

shielded metal arc welding

TGSCC:

transgranular stress corrosion cracking

UT:

ultrasonic testing

1.2. Introduction with general description of the systems

1.2.1. Pressurized water reactor (Kaercher 2002; Cattant 2014; Hutin 2017a)

From the end of the 1960s, light water reactors (LWRs) have been used extensively in many countries around the world for electricity production.

The predominant reactor design worldwide is the pressurized water reactor (PWR), accounting for about two-thirds of the installed capacity, followed by boiling water reactors (BWRs) at about 20%.

Figure 1.1 shows an overview of a typical modern PWR. The condenser is generally cooled by sea, brackish or river water. However, in the latter situation, when the river flow is too low, cooling towers are used, as shown in the picture.

Figure 1.1.Sketch of a PWR presenting the main buildings and facilities

(courtesy of EDF).

The flow of primary water is maintained by the pressurizer and flows through the reactor coolant system (RCS) by reactor coolant pumps. The typical RCS water conditions are as follows:

high temperature: between 285°C (cold leg, reactor pressure vessel inlet) and 325°C (hot leg, reactor pressure vessel outlet);

high pressure: around 155 bars;

reducing environment for water radiolysis mitigation by hydrogen injection (typically: 30 cc/kg);

presence of boric acid for the neutronic reaction control. The primary water pH

300°C

is typically adjusted to 7.2 by way of lithium hydroxide injection (a few ppm maximum).

The RCS heat energy is exchanged with the secondary circuit through the steam generator tubes. The secondary water is heated to steam that powers the high- and low-pressure turbines, which turn the main generator for electricity production.

The steam is then condensed to water and flows back to the steam generators. For thermodynamic considerations (cycle efficiency), the water is steam-heated before entering the steam generator and the steam is dried and reheated between the high- and low-pressure turbines in the moisture separator-reheaters (MSRs). At the steam generator outlet, the typical steam conditions are as follows:

temperature: 286°C;

pressure: 70 bars.

A schematic of the primary and secondary circuits and materials of construction is presented in Figures 1.2 and 1.3 for PWR and BWR, respectively.

There are hundreds of various systems in a PWR; some related to the primary circuit are indicated.

Although optimized, the water and steam environments are still harsh enough to pose a threat to the systems materials, at times leading to corrosion failures. Mechanical loads can also result in mechanical failures, sometimes in conjunction with corrosion, like in the cases of stress corrosion cracking (SCC) and corrosion fatigue.

Figure 1.2.Principle of PWR: main components and materials. In PWR, the water heated in the RPV is used to heat a secondary circuit.

Figure 1.3.Principle of BWR: main components and materials. In BWR, the water heated in the RPV directly feeds the turbine.

1.2.2. BWR (Boiron 2011; Cattant 2014)

Figure 1.4 presents the BWR power cycle. Basically, the BWR has a lower reactor pressure vessel (RPV) pressure and a simplified steam cycle compared to PWR. The RPV pressure is around 7 MPa (1,020 psig). The temperature in the RPV reaches 288°C (550°F).

As the steam is generated in the RPV, bulk boiling is allowed in the BWR core, which involves limitations regarding possible water/steam chemical conditioning (i.e. no boric acid or lithium hydroxide injection).

One particular component found in BWRs is jet pumps, which provide flow to control reactor power, which enables a higher power level without increasing the RPV size. Jet pumps also provide part of the boundary required to maintain two-thirds of the core height following a recirculation line break event.

Figure 1.4.Advanced BWR power cycle

(Cattant (2014), courtesy of GE).

Figure 1.5 presents a typical BWR lower plenum, with the presence of:

control rod drive guide tubes;

control rod blades;

control rod drive housings;

stub tubes;

in-core housings;

guide tubes;

flux monitor dry tubes.

Another major component of BWR core internals is the core shroud, which is a large austenitic stainless steel cylinder surrounding the core. It separates upward flow through the core from downward flow in the downcomer annulus. It also provides a two-thirds core height floodable volume.

Another big piece of equipment sitting at the top of BWR RPV internals is the steam dryer. The steam dryer provides a 99.9% steam flow to the main turbine. In the dryer, wet steam is forced through the dryer panels horizontally:

the steam is forced to make a series of rapid changes in direction;

the moisture is thrown to the outside.

Note that initial power uprate plants experienced flow induced vibrations, which have been minimized by design improvements.

Figure 1.5.BWR typical lower plenum

(Cattant (2014), courtesy of GE).

1.3. Requirements for materials (Meyzaud and Vieillard-Baron 1998; Champigny 2005; Lemaignan 2010; Zinkle and Was 2013)

The selection of materials must take into account some drivers that are specific to LWR, as summarized below:

large size components (with homogeneous properties), such as the reactor vessel, the primary pump, the pressurizer or the steam generators;

good corrosion resistance (primary and secondary circuits);

knowledge of materials in-service behavior, in accordance with regulator requirements concerning fatigue, irradiation, thermal aging, corrosion, etc.;

prevention of the risk of brittle fracture;

limitation of activation and contamination of materials (avoiding Co, etc.).

In order to reach the required quality level, nuclear construction codes (ASME, RCC-M, etc.) require:

reduced chemical composition ranges for main components and strict control of residual elements;

non-destructive examinations at various stages of manufacturing;

technical qualification for the first manufactured component (additional testing, technical qualification);

identification of main parameters controlling material properties, in technical manufacturing programs of suppliers.

1.3.1. Brittle fracture

One of the most significant regulatory requirements is the prevention of the risk of brittle fracture, defined as a sudden fracture, without any preliminary significant overall deformation. As far as the material is concerned, brittle fracture prevention is controlled through dedicated requirements for notch impact toughness (Charpy tests) at several temperatures, at least down to 0°C, and ductility (i.e. elongation at rupture and reduction of area through tensile tests).

To prevent any risk of brittle fracture, strict requirements are imposed on materials at the procurement stage. As an example, for heavy components, additional margins have been introduced in the equipment specification by imposing a maximum reference Nihil Ductility Temperature (RTNDT) value equal to −20°C. Nevertheless, this is not sufficient in cases where aging may reduce the resistance to brittle fracture through either an increase of the brittle-ductile transition temperature or an intrinsic reduction of fracture toughness. Further precautions have been taken to limit the effects of aging, as described below.

1.3.2. Corrosion

Materials corrosion is also an important issue not only from the viewpoint of the damage to the component itself, but also due to the generation of corrosion products. The materials must provide a high degree of protection against general corrosion, as well as the different forms of localized corrosion such as stress corrosion cracking, pitting or crevice corrosion.

The undesirable consequences of the migration of corrosion products may also be significant. Deposition of such products on fuel elements and heat exchanger surfaces may lead to ineffective heat transfers and the formation of radioactive isotopes after activation inside the reactor core.

Strong efforts have been applied to limit the cobalt contents of materials, with the aim to reduce activation, especially in steam generator tubes, for which an upper limit of 0.015% Co has been imposed. Besides, for main primary component surfaces directly in contact with the primary coolant, an upper limit of 0.060% Co has been set in stainless steels. The prevention of potential cobalt release has also lead to efforts to develop alternative materials to replace cobalt base alloys for hard facing. When available, these alternative materials can be used, particularly in valves.

The selection of materials for new reactors has resulted from choices made in the past for French and German reactors (N4 and Konvoi, respectively). The specific requirements for these materials take field experience and improvement in technical knowledge into account. They are de facto more severe now than they used to be for original units.

The materials used in nuclear power plants (NPPs) were naturally chosen from the experience of fossil power plants; it was indeed essential to rely on a set of reliable and very long-lasting data in order to focus innovation on a few specificities related to the operating conditions of nuclear reactors:

effect of neutron bombardment on reactor pressure vessel materials or vessel internal structures;

new nature of certain chemical environment (boric water of the primary water or steam that is very charged with moisture in turbines, etc.);

difficulty of intervention for repair in radioactive environment.

The materials used are mostly steels. The logic is to consider the most usual steels (with the lowest cost and whose conditions of use are the largest) in the first place and to only choose more sophisticated steels or alloys to meet particular constraints of use.

Non-alloy ferritic steels are the most conventional boiler construction steels. These steels are suitable for most pipes of water-steam circuits, outside the creep range. However, the use of other steels or alloys is necessary when higher mechanical characteristics, improved corrosion resistance or special physical properties are required.

Low alloy steels are used when higher mechanical characteristics are needed in the field of temperatures below 350°C, for example, for pressure capacities of a large diameter (concerning limit thicknesses): case of the reactor pressure vessel. However, these materials have several disadvantages: the increase in tensile characteristics is generally to the detriment of brittle fracture resistance (resilience, toughness); they are sensitive to hydrogen embrittlement, which can be connected to corrosion or welding operations; finally, the quench rate varies greatly between the skin and the core during the manufacture of thick parts, which could involve detrimental heterogeneities of structure and properties.

Austenitic stainless steel grades selected for PWRs are usually “very low carbon” grades or stabilized (i.e. containing titanium or niobium) grades to prevent sensitization during manufacturing operations. Most often, considering corrosion susceptibility leads to the choice of stainless steels (addition of chromium with contents greater than 12–13%). This choice may be necessary in primary circuits where the pH cannot be set in the range minimizing general corrosion of carbon steels. The selection of stainless steels helps to avoid a high potential corrosion rate and limit the harmful consequences of an excessive release of corrosion products in the circuit, which could be activated in the irradiated areas. Although highly resistant to general corrosion, stainless steels can be susceptible to various localized corrosion phenomena such as intergranular corrosion, stress corrosion cracking or pitting corrosion.

Among stainless steels, martensitic stainless steels are used for their high mechanical characteristics, which can be adjustable by heat treatment (bolting, valve shaft, etc.). However, those steels, in particular the high chromium grades, may undergo thermal aging and embrittlement at reactor operating temperatures (300–350°C), which may strongly limit their practical use.

Austenitic stainless steels have, for the basic grades, low tensile characteristics, but good weldability and easy formability. They are widely used for pipes subject to severe corrosion conditions (primary circuit piping, etc.). For castings and welding filler metals, austenitic-ferritic grades are selected, as a minimum amount of ferrite prevents hot cracking during solidification. However, ferrite is sensitive to thermal aging at working temperatures and its content must be limited.

Ferritic stainless steels are sometimes selected in very singular cases.

In the case of the specific corrosion conditions encountered by steam generators, tubes and other components (stress corrosion expected), nickel-based alloys are selected.

All materials must provide a high degree of protection against general corrosion and the various forms of localized corrosion, such as stress corrosion cracking, pitting or crevice corrosion. Indeed, the migration and deposition of corrosion products may lead to ineffective heat transfers and the formation of radioactive isotopes after activation inside the reactor core.

Zirconium alloys were selected for the fuel cladding based on their neutron transparency and good resistance to corrosion in water, and when in contact with the fuel rods. Discussing materials and processes for the fuel assembly is beyond the scope of this chapter.

The main difference between PWRs and BWRs is that the latter consist of single water circuits, which are designed to allow boiling to occur in the core, with steam flowing directly to the turbine, which eliminates the steam generator and pressurizer found in PWRs. Operating temperatures remain close for reactors (lower for BWRs), with similar stress and radiation environments (lower for BWRs), and with the specificity of a more oxygenated environment for BWRs. Consequently, most of the structural alloys are very similar in both reactor types.

1.4. Material choices and basic properties

All of the considerations on materials requirements lead to the choice of commercial materials grades, which are well known by manufacturers and easy to produce, such as carbon steels, low alloy steels (LASs), austenitic stainless steels and nickel base alloys. Concrete (e.g. for containment building and cooling towers), polymers (e.g. for cables, composite materials and HDPE) and miscellaneous materials (e.g. hard-facing materials (stellite), copper and titanium base (condenser tubes), high strength materials (bolting)) are briefly considered at the end of this chapter.

Table 1.1.Main basic properties and application of the different classes of used materials in an LWR

Carbon steels

Low-alloy steel

Austenitic stainless steel

Ni alloys

Cristal structure

Body centered cubic (cristallography) (BCC)

BCC

Face centered cubic (cristallography) (FCC)

FCC

Microstructure

Ferritic, bainitic

Ferritic, bainitic

Austenitic

Austenitic

Main alloying elements

0.05–0.35% C, 0.5–1.5% Mn, Si, etc. Total (w/o Mn) < 1%

C < 0.27% Mn, Ni, Mo, Cr, etc. Total < 5%

~18% Cr, ~10% Ni Mn < 2%, Si < 1% C < 0.08%

~15% Cr, ~10% Fe ~30% Cr, ~10% Fe 0.04–0.08% C

Main grades

TU, A-48; TU, A-42, P355 NH A106, A333, A515, etc.

16MND5, 20MND5 A533 Cl.1, A508 Cl.2/3, 15Kh2MFA, 15Kh2NMFA

304 L, 316 (L), 321, 347 Cast: CF3M, CF8M, etc. A800

Alloy 600 Alloy 690, others; X750, 718, etc.

Filler products

A42, A48, SA105, SA216, SA350, etc.

SA508 Gr 2/3, SA533 Type A/B

308L, 309L

182/82, 152/52

Fabrication

Forged, rolled or cast

Forged or rolled, cladded

Forged, rolled or cast

Forged, rolled

Heat treatment

Austenitizing, air cooling

Quenched + tempered + post-weld HT

Solution annealed (SA) (+ cold work)

Mill annealed Thermally treated

Weldability

Cold cracking sensitivity

Cold cracking sensitivity

Hot cracking sensitivity

Hot cracking sensitivity

Yield stress

250–450 MPa

250–450 MPa

~200 MPa (SA) if cold worked (CW), YS increases

~300 MPa (SA) if CW, YS increase

Toughness/ductility

High ductile brittle transition temperature (DBTT)

High, DBTT

Very high, no DBTT

Very high, no DBTT

Irradiation resistance

N/A

Moderate (% Cu, P low)

High

High

Thermal aging

Moderate (% P)

Moderate (% P)

CF3M, CF8M

N/A

Corrosion resistance

Risk for flow-assisted corrosion (FAC) High resistance SCC

Moderate

High

High

Main components

Secondary piping, tubes and plates for exchangers

Vessels (RPV, pressurizer, steam generator)

Primary piping, surge line, RPV internals, pump casing, etc.

Steam generator (SG) tubing, divider plate, BMI, Head penetration, etc.

1.4.1. Carbon and LASs (SFEN 2008; Cattant 2014; Gras 2017)

Carbon and LASs have been selected for LWR applications, mainly because of the following properties:

rather low cost;

good mechanical properties;

good weldability;

high SCC resistance;

rather low neutron embrittlement;

good thermal conductivity (tubing applications).

Carbon steels and LASs can be supplied as:

black and hot-dipped galvanized steel pipe;

carbon steel forgings for piping applications;

seamless carbon steel pipe for high-temperature service;

pipe, steel, electric-fusion (arc)-welded (sizes NPS 16 and over);

steel castings, carbon steels, suitable for fusion welding for high-temperature service;

pipe fittings of wrought carbon steel and LAS for moderate and elevated temperature service;

low and intermediate tensile strength carbon steel plates;

seamless and welded steel pipe for low-temperature service;

pipe fittings of wrought carbon steel and LAS for low-temperature service;

tubes for heat exchangers;

pressure vessel plates, carbon steel, for intermediate-temperature and higher temperature service;

pressure vessel plates, carbon steel, for moderate-temperature and lower temperature service.

1.4.2. Austenitic stainless steels (SFEN 2008; Cattant 2014; Hutin 2017b)

Austenitic stainless steels have been selected in LWRs because of the following properties:

corrosion resistant to the surfaces directly in contact with the primary coolant;

good weldability;

high mechanical properties when cold worked;

rather easy casting with some amount of ferrite.

Austenitic stainless steel grades are now “very low carbon” grades or stabilized (i.e. containing titanium or niobium) grades to prevent intergranular corrosion sensitization.

For primary components with surfaces directly in contact with the primary coolant, an upper limit of 0.060% Co has been set in stainless steels.

Austenitic stainless steels can be supplied as:

forged: valves, fittings, pipe flanges;

rolled: plates, sheets, tubes;

castings: elbows, pump casings, pipes;

tubes (also: ferritic stainless steels).

Alloy 800 is also considered as an austenitic stainless steel, mainly used for steam generator tubing.

1.4.3. Nickel base alloys (SFEN 2008; Cattant 2014)

Nickel-base alloys have been selected for LWR applications, mainly because of the following properties:

corrosion resistance (Cl

-

, general, primary water stress corrosion cracking (PWSCC));

high mechanical properties;

expansion coefficient close to those of LASs (interesting for heat exchanger tubing).

To reduce activation, the cobalt content was limited to 0.015% Co, for instance, in steam generator tubes.

After the first evidence of the SCC susceptibility of former alloy 600 exposed to primary PWR environment as long as secondary corrosion and SCC, improvements for steam generator tubes have been put in place: first with a dedicated heat treatment at 715°C, and finally with the choice of 690 TT alloy with higher chromium content.

Nickel-base alloys can be supplied as:

rods, bars and wires;

seamless pipes and tubes;

plates, sheets and strips;

forgings.

1.4.4. Other materials

1.4.4.1. Hardfacing materials

Hard-facing alloys have been selected for LWR applications, mainly because of the following properties:

high wear resistance;

high mechanical properties;

good corrosion resistance;

good thermal shock resistance.

Potential cobalt release also triggers efforts to develop alternative materials to replace cobalt base alloys for hard facing (Norem).

1.4.4.2. Condenser tubes (Gras 2017)

In the 1970's, the construction of the tube bundles of PWR condensers was based, for power plants cooled by river water, on conventional technology, with brass tubes (copper-zinc alloy), inherited from the experience gained from fossil power plants. These condensers quickly presented generic problems that past experience did not predict:

erosion by wet steam of impact tubes and peripherals;

vibration of the tubes, resulting in fatigue ruptures or contact wear;

ammoniacal corrosion of the brass tubes of the central part of the beams where the non-condensables are concentrated;

stress corrosion in the rolling zone that can be initiated from the inside of the tubes (refrigeration water side) or from the outside (ammonia-conditioned steam side);

flow-assisted corrosion (FAC) on the raw water side that results in a regular thinning of the inner skin tube at a speed of about 50 μm/year, affecting the entire bundle of tubes and determining the life of the condenser (about 10 years).

These problems led to the adoption of AISI 316L austenitic stainless steel for the rehabilitation of damaged condensers and the construction of new condensers. It should be noted that the gradual replacement of copper alloys, a bacteriostatic material (i.e. preventing the creation of a biofilm), by austenitic stainless steel, has given rise to a new concern: the appearance of pathogenic amoebae in the environment in summer.

For power plants cooled by sea water, the construction was based on the choice of titanium. The results of the first years of operation showed that the technological choices made for condensers cooled by sea water were overall satisfactory. These devices do not pose major problems for the operator and are not expected to undergo major changes in the future, as long as the cathodic protection is properly adjusted and monitored over time.

1.4.4.3. High-strength alloys

The typical high-strength alloys used in NPPs are as follows: 17-4 PH; X750; X718; A286; cold worked (CW) austenitic stainless steels; martensitic stainless steels; Ni-base alloys.

These alloys have been selected for LWR applications, mainly because of the following properties:

high mechanical properties;

quite often: good corrosion resistance.

High-strength alloys can be supplied as (not all products for all alloys) bars, plates, sheets, strips, wires, forgings and forging billets, seamless tubes, rings, turbine blades and castings.

The attachment of the closer head to the reactor pressure vessel is carried out using studs screwed into the main flange on one side and through the lid flange on the other, so as to ensure proper tightening by nuts placed above the lid. These studs are machined in forged bars made of low-alloy steel, containing nickel, chromium, molybdenum and vanadium to give the metal the necessary high tensile characteristics, especially in temperature, maintaining good resistance to fast fracture.

The bolting of reactor coolant pumps, whose diameter is much smaller, is carried out using forged steel bars containing nickel, chromium, molybdenum, and those of steam generators and pressurizers are made using forged steel or rolled with chromium and molybdenum with a possible addition of vanadium. In each case, the codes (RCC-M, etc.) provide precise recipe acceptance conditions regarding the mechanical characteristics and non-destructive tests to be carried out.

1.4.4.4. Concrete (Bouniol 2001)

Concrete and, more generally, cementitious materials are used in the civil engineering works of NPPs (reactor building, cooling tower, etc.). The aging mechanisms of cementitious materials are numerous due to their heterogeneous nature and porous structure. Thus, because of this porosity, the species present in the external environment are likely to penetrate the material, which can cause a depassivation of steel in reinforced concrete structures. In particular, carbon dioxide from the atmosphere causes a degradation of concrete, called carbonation, which modifies its physical and chemical properties, and, in particular, leads to the pH of the interstitial solution falling to values of about 8. Another common cause of concrete degradation comes from the diffusion of chlorides. The stages of corrosion and the different stages of the evolution of the degradation of a reinforced concrete structure can be summarized as follows:

a transport step of aggressive agents (mainly carbon dioxide and chlorides; it could also be sulfates) from water and oxygen;

a corrosion initiation step (depassivation of reinforcements). The two main phenomena at the origin of the depassivation of steel in cementitious medium are the decrease in the pH of the interstitial solution due to carbonation and the penetration of chloride ions;

a stage of rust growth (in an airy environment), which gradually extends over the steel. The resulting volume expansion leads to damage to the concrete (chips, splinters and cracks).

1.4.4.5. Polymers

Polymers, in the broad sense of the term (elastomers, thermoplastics, composite matrix, ion exchange resins, etc.), are widely used in NPPs. The mechanisms of aging depend on the material and the conditions of use or exposure. Physical aging is often distinguished from chemical aging, although in reality both phenomena interfere. Irradiation aging combined with thermal aging is also possible for cables.

1.5. Fabrication and joining

Nuclear power imposes particular developments and manufacturing requirements:

The large sizes of some of the forged pieces (pressure vessel shells, closure head, etc.) reach the practical limits of feasibility. It is then necessary to control the composition of very large ingots and their homogeneity. Macro-segregation must be minimized during the solidification of ingots (hence the use of hollow ingots). Finally, it is necessary to have presses adapted to the size of these components.

The composition requirements/specifications impose full control of the metallurgy, with many preliminary tests and validation steps in the reactor, which require examinations in “hot laboratories”.

1.5.1. Forging of Ferritic components: from ingot to component (Vieillard-Baron and Meyzaud 1998; Destre and Izard 2015)

Pressure vessels of a large diameter (with concern to limit thicknesses) – reactor pressure vessels, steam generator (SG) parts, tube sheets and presssurizer shells – are generally forged from LAS ingots.

Since the 1960s, the various changes in the grade of steel in PWRs reflect the improvement in the steelmakers’ processes (reduction in the maximum values of impurities as S and P, narrowing of the variation ranges of certain elements, etc.), and a better knowledge of the effect of additives elements on the properties of steels.

Advances in steelmaking techniques have been very beneficial for these steels. In particular, from the 1980s, the techniques of vacuum degassing and ladle refining have made it possible to obtain:

better analytical precision;

very low contents of elements such as hydrogen (harmful for cold cracking), sulfur (responsible for the anisotropy of the characteristics and a reduction in the level of the ductile bearing) and phosphorus (influence on the fragility tempering, resistance to irradiation and thermal aging);

better control of the aluminum content for the trapping of free nitrogen by formation of AlN;

minimization of metal inclusions, endogenous or exogenous.

Whatever their geometry, components intended for the nuclear industry must satisfy increasingly demanding design specifications. Among all of the criteria required, we can cite those relating to the inevitable phenomena of segregation that particularly concern forged parts from large ingots:

limitation of the rate of macro-segregation remaining;

absence or limitation of area with “ghost lines” (segregation lines) on faces to be coated by welding.

The heterogeneity of large forge ingots, induced by solutal and thermal convection currents, and exacerbated by long solidification times, is well known (Figure 1.6).

Figure 1.6.Phenomena of segregation in large ingots: improvement of numerical simulation.

Although for most hollow parts, most of the macro-segregation is removed by hot drilling during forging, it is impossible to remove the entire area of “ghost lines”. Gradually, over time, the approach was therefore directed toward adapting the types of ingots according to the parts to be delivered, in order to reduce the level of macro-segregation and minimize the disadvantages linked to segregation and “ghost lines”. This was done by locating them in parts where they are less detrimental, that is, within the thickness. For this, different types of ingots were developed at the end of the 1970s, such as the directional solidification ingot and the hollow ingot. Nevertheless, for very thick components such as SG tube sheets, it is still not possible to eliminate all of the heterogeneities.

The forging of LAS is carried out between 1,300 and 1,000°C in several operations, such as blooming, crushing, hot drilling, drawing on a mandrel and horning, in order to obtain a deformation rate, allowing the desired properties to be obtained. The forging ranges are optimized regularly to obtain parts as close as possible to finished parts, to eliminate welds, or to move them to more accessible areas. For example, we can cite the manufacture of steam generator heads with pipes extruded by hot punching, and more recently, the development of forging the vessel head of 900 MW monobloc reactors, which makes it possible to eliminate the flange-cap weld.

1.5.2. Casting… to forging (Vieillard-Baron and Meyzaud 1998)

A major part of the components of the main coolant lines are made from forged austenitic stainless steels tubes.

But, historically, for primary pump casing, elbows, fittings, valves, etc., cast duplex austenitic-ferritic stainless steels were mostly used. As the production progressed, the sizing of the coolers and risers made it possible to limit the appearance of shrinkage cavities during the solidification of these cast components. These steels have a ferrite content ranging from about 10 to 25% to avoid solidification cracking and obtain higher tensile properties. At operating temperature however, due to thermal aging, hardening of the ferrite phase takes place, inducing embrittlement.

A molybdenum-free grade (CF3 type), less susceptible to aging, has been selected; more recently, cast duplex steels in primary pipe branches have been replaced by forged austenitic grades that are not susceptible to thermal aging. Then, for the recent fabrications, mostly the pump casings are made from cast duplex steel.

1.5.3. Forging of PWR SG tube bundle

The U-tube bundle transfers heat from the primary system to the secondary system. The ends of the U-tubes are inserted in a tube sheet 500–600 mm thick. Their entire inserted lengths are roll-expanded and the tube-ends are then welded to the tube-sheet. Straight leg portions of the tube bundle are held in place by tube support plates and bent portions by anti-vibration bars.

Tube blanks are made from billets obtained by hot forging of ingots. The billets are cut, machined and extruded, then converted by drawing or rolling into blanks. The blanks are then cold worked by drawing or rolling in several passes, with intermediate annealing. They undergo final annealing at a specified temperature, then are subjected to a supplementary 715 ± 15°C heat treatment, followed by bending. U-bends with radii less than 10 times tube diameter are subjected to an additional stress-relieving heat treatment for 2 h at 715°C.

1.5.4. Welding and cladding (Buisine and Perrat 201)

1.5.4.1. Welding

The main welding techniques used are arc welding with filler material for all materials (submerged arc welding (SAW), shielded metal arc welding (SMAW) and gas tungsten arc welding (GTAW)). For the record, electroslag welding was formerly used for pump casing welding.

Table 1.2.Main welding techniques used

Component/materials

Welding technique

Source

Protection

RPV, SG, PZR, etc. (

LAS

) MCL/

austenitic stainless steel

Shielded metal arc welding (SMAW)

Arc

Slag

RPV, SG, PZR, etc. (

LAS)

MCL/

austenitic stainless steel

Submerged arc welding (SAW)

Arc

Slag

MCL/

austenitic stainless steel

Gas tungsten arc welding (GTAW)

Arc

Gas

Pump casing parts

Duplex stainless steel

Electroslag welding (ESW)

Joule effect

Slag

1.5.4.2. Cladding

The inner wall of the vessel is coated with one or two layers of austenitic stainless steel, which protects it from corrosion by the cooling water. To avoid intergranular corrosion, the number of layers and the mode of deposition are defined so that the carbon content of the inner layer is at most equal to 0.035%, hence the need to produce a two-layer coating.

Two deposition processes were used: the conventional flux arc welding process with strip electrode (strip width 60 or 90 mm), then the electroconductive slag welding process with strip electrode (so-called Maglay process, strip width 75 or 150 mm).

1.5.5. Heat treatment (Vieillard-Baron and Meyzaud 1998; Destre and Yzard 2015)

After forging, LAS components undergo normalization heat treatment and tempering at around 650°C to facilitate subsequent ultrasonic inspection operations in the semi-finished state.

After roughing and inspection, the last quality heat treatment includes (i) austenitization at 875°C, which, among other things, makes it possible to refine the grain, (ii) quenching by immersion in water, which provides the required bainitic microstructure and (iii) tempering around 650°C, followed by slow cooling.

After welding, relaxation heat treatment is required for unalloyed or low alloyed ferritic steels around 600°C, but is not required (because it is potentially unfavorable) for austenitic stainless steels (due to potential risk of sensitization or formation of embrittling phases).

Austenitic stainless steels are generally only solution annealed at 1,050°C and water cooled.

1.5.6. Manufacturing defects (links with fabrication)

Among the different types of welding defects that can affect the components (geometrical defects, technological defects and metallurgical defects), a few of them are more specific to LWR components: for example, underclad defects.

1.5.6.1. Underclad defects

Some weld procedures used for stainless steel cladding of pressure vessels (reactor, steam generator and pressurizer) resulted in manufacturing defects in the heat affected zone of the base metal under the cladding.

Two types of underclad cracking have been identified: reheat cracking and cold cracking.

Reheat cracking occurred as a result of post weld heat treatment of austenitic stainless steel cladding applied using the high-heat-input welding process. Reheat cracking was attributed to high residual stresses near the yield strength in the weld metal and base metal interface, combined with a sensitive microstructure of the base metal: coarse grains and segregated lines. The cracks are in the base metal region directly beneath the cladding and are confined to a region approximately 0.125 inches deep and 0.4 inches long. This problem was eliminated by modifying the welding sequence of the second coating layer in the affected areas.

Cold cracking occurred after deposition of the second and third layers of cladding, where no pre-heating or post-heating was applied during the cladding procedure. The cold cracking was attributed to residual stresses near the yield strength in the weld metal and base metal interface after cladding deposition, combined with a crack-sensitive microstructure in the heat-affected zone (HAZ) and a high level of diffusible hydrogen in the austenitic stainless steel. The hydrogen diffused into the HAZ and caused cold (hydrogen-induced) cracking as the HAZ cooled. Corrective measures consisted of heating the parts to more than 150°C in the first and second layers of coating. Another measure consisted of ensuring that the segregations were not present on the wall to be coated using a hollow ingot.

Other manufacturing defects such as hydrogen flakes sometimes affect LAS components due to insufficient degassing treatments. Welding defects as hot cracking can also affect austenitic stainless steels. Cast components are affected by solidification shrinkage defects.

1.5.7. Non-destructive testing (links with expected defects) (Champigny 2005)

Non-destructive techniques are specifically developed and used to detect the most current defects that can affect the reactor components.

According to the type and location of the defect to be detected, the techniques presented in Table 1.3 are used.

Table 1.3.